Pacific Northwest Laboratory conducted experiments with aquifer sediments and leaching solution (lixiviant) from an in-situleachuranium mine. The data from these laboratory experiments and information on the normal distribution of elements associated with roll-front uranium deposits provide evidence that natural processes can enhance restoration of aquifers affected by leach mining. Our experiments show that the concentration of uranium (U) in solution can decrease at least an order of magnitude (from 50 to less than 5 ppM U) due to reactions between the lixiviant and sediment, and that a uranium solid, possibly amorphous uranium dioxide, (UO/sub 2/), can limit the concentration of uranium in a solution in contact with reduced sediment. The concentrations of As, Se, and Mo in an oxidizing lixiviant should also decrease as a result of redox and precipitation reactions between the solution and sediment. The lixiviant concentrations of major anions (chloride and sulfate) other than carbonate were not affected by short-term (less than one week) contact with the aquifer sediments. This is also true of the total dissolved solids level of the solution. Consequently, we recommend that these solution parameters be used as indicators of an excursion of leaching solution from the leach field. Our experiments have shown that natural aquifer processes can affect the solution concentration of certain constituents. This effect should be considered when guidelines for aquifer restoration are established.

In the last few years, there has been a significant increase in the demand for Uranium as historical inventories have been consumed and new reactor orders are being placed. Numerous mineralized properties around the world are being evaluated for Uranium recovery and new mining / milling projects are being evaluated and developed. Ore bodies which are considered uneconomical to mine by conventional methods such as tunneling or open pits, can be candidates for non-conventional recovery techniques, involving considerably less capital expenditure. Technologies such as UraniumInSituLeaching / In Situ Recovery (ISL / ISR - also referred to as 'solution mining'), have enabled commercial scale mining and milling of relatively small ore pockets of lower grade, and are expected to make a significant contribution to overall world wide uranium supplies over the next ten years. Commercial size solution mining production facilities have operated in the US since the mid 1970's. However, current designs are expected to result in less radiological wastes and emissions relative to these 'first' generation plants (which were designed, constructed and operated through the 1980's). These early designs typically used alkaline leach chemistries in situ including use of ammonium carbonate which resulted in groundwater restoration challenges, open to air recovery vessels and high temperature calcining systems for final product drying vs the 'zero emissions' vacuum dryers as typically used today. Improved containment, automation and instrumentation control and use of vacuum dryers in the design of current generation plants are expected to reduce production of secondary waste byproduct material, reduce Radon emissions and reduce potential for employee exposure to uranium concentrate aerosols at the back end of the milling process. In Situ Recovery in the U.S. typically involves the circulation of groundwater, fortified with oxidizing (gaseous oxygen e.g) and complexing agents (carbon dioxide, e.g) into an ore body, solubilizing the uranium in situ, and then pumping the solutions to the surface where they are fed to a processing plant ( mill). Processing involves ion exchange and may also include precipitation, drying or calcining and packaging operations depending on facility specifics. This paper presents an overview of the ISR process and the health physics monitoring programs developed at a number of commercial scale ISL / ISR Uranium recovery and production facilities as a result of the radiological character of these processes. Although many radiological aspects of the process are similar to that of conventional mills, conventional-type tailings as such are not generated. However, liquid and solid byproduct materials may be generated and impounded. The quantity and radiological character of these by products are related to facility specifics. Some special monitoring considerations are presented which are required due to the manner in which radon gas is evolved in the process and the unique aspects of controlling solution flow patterns underground. The radiological character of these processes are described using empirical data collected from many operating facilities. Additionally, the major aspects of the health physics and radiation protection programs that were developed at these first generation facilities are discussed and contrasted to circumstances of the current generation and state of the art of uranium ISR technologies and facilities. In summary: This paper has presented an overview of in situ Uranium recovery processes and associated major radiological aspects and monitoring considerations. Admittedly, the purpose was to present an overview of those special health physics considerations dictated by the in situ Uranium recovery technology, to point out similarities and differences to conventional mill programs and to contrast these alkaline leach facilities to modern day ISR designs. As evidenced by the large number of ISR projects currently under development in the U.S. and worldwide, non conventional Uranium recovery techniques

A floating plant has been designed to extract uranium from seawater using solid adsorbents. Ore is removed from the adsorbent material by means of a solvent and concentrated in ion exchangers. Seawater is supplied to the adsorbent inside by wave energy and is based on the principle that waves will rush up a sloping plane that is partly submerged and fill a reservoir to a level higher than the still water level in the sea. The company projects that an offshore plant for recovering 600 tons of uranium/yr would comprise 22 floating concrete units, each measuring 430 x 75 meters.

Safeguarding sensitive fuel cycle technology such as uranium enrichment is a critical component in preventing the spread of nuclear weapons. A useful tool for the nuclear materials accountancy of such a plant would be an instrument that measured the uranium content of UF{sub 6} cylinders. The Uranium Cylinder Assay System (UCAS) was designed for Japan Nuclear Fuel Limited (JNFL) for use in the Rokkasho Enrichment Plant in Japan for this purpose. It uses total neutron counting to determine uranium mass in UF{sub 6} cylinders given a known enrichment. This paper describes the design of UCAS, which includes features to allow for unattended operation. It can be used on 30B and 48Y cylinders to measure depleted, natural, and enriched uranium. It can also be used to assess the amount of uranium in decommissioned equipment and waste containers. Experimental measurements have been carried out in the laboratory and these are in good agreement with the Monte Carlo modeling results.

The decision to cease all US Department of Energy (DOE) reprocessing of nuclear fuels was made on April 28, 1992. This study provides insight into and a comparison of the management, technical, compliance, and safety strategies for deactivating the Idaho Chemical Processing Plant (ICPP) at Westinghouse Idaho Nuclear Company (WINCO) and the Westinghouse Hanford Company (WHC) Plutonium-Uranium Extraction (PUREX) Plant. The purpose of this study is to ensure that lessons-learned and future plans are coordinated between the two facilities.

A proposed solid waste landfill at Paducah Gaseous Diffusion Plant (PGDP) will accept wastes generated during normal plant operations that are considered to be non-radioactive. However, nearly all solid waste from any source or facility contains small amounts of radioactive material, due to the presence in most materials of trace quantities of such naturally occurring radionuclides as uranium and thorium. This paper describes an evaluation of operating limits, which are protective of public health and the environment, that would allow waste materials containing small amounts of radioactive material to be sent to a new solid waste landfill at PGDP. The operating limits are expressed as limits on concentrations of radionuclides in waste materials that could be sent to the landfill based on a site-specific analysis of the performance of the facility. These limits are advantageous to PGDP and DOE for several reasons. Most importantly, substantial cost savings in the management of waste is achieved. In addition, certain liabilities that could result from shipment of wastes to a commercial off-site solid waste landfill are avoided. Finally, assurance that disposal operations at the PGDP landfill are protective of public health and the environment is provided by establishing verifiable operating limits for small amounts of radioactive material; rather than relying solely on administrative controls. The operating limit determined in this study has been presented to the Commonwealth of Kentucky and accepted as a condition to be attached to the operating permit for the solid waste landfill.

Isotopically depleted UF{sub 6} (uranium hexafluoride) accumulates at a rate five to ten times greater than the enriched product and is stored in steel vessels at the enrichment plant sites. There are approximately 55,000 large cylinders now in storage at Paducah, Kentucky; Portsmouth, Ohio; and Oak Ridge, Tennessee. Most of them contain a nominal 14 tons of depleted UF{sub 6}. Some of these cylinders have been in the unprotected outdoor storage environment for periods approaching 40 years. Storage experience, supplemented by limited corrosion data, suggests a service life of about 70 years under optimum conditions for the 48-in. diameter, 5/16-in.-wall pressure vessels (100 psi working pressure), using a conservative industry-established 1/4-in.-wall thickness as the service limit. In the past few years, however, factors other than atmospheric corrosion have become apparent that adversely affect the serviceability of small numbers of the storage containers and that indicate the need for a managed program to ensure maintenance ofcontainment integrity for all the cylinders in storage. The program includes periodic visual inspections of cylinders and storage yards with documentation for comparison with other inspections, a group of corrosion test programs to permit cylinder life forecasts, and identification of (and scheduling for remedial action) situations in which defects, due to handling damage or accelerated corrosion, can seriously shorten the storage life or compromise the containment integrity of individual cylinders. The program also includes rupture testing to assess the effects of certain classes of damage on overall cylinder strength, aswell as ongoing reviews of specifications, procedures, practices, and inspection results to effect improvements in handling safety, containment integrity, and storage life.

The partition ratio of cerium (Ce) and uranium (U) in the ingot, slag and dust phases has been investigated for the effect of the slag type, slag concentration and basicity in an electric arc melting process. An electric arc furnace (EAF) was used to melt the stainless steel wastes, simulated by uranium oxide and the real wastes from the uranium conversion plant in Korea Atomic Energy Research Institute (KAERI). The composition of the slag former used to capture the contaminants such as uranium, cerium, and cesium during the melt decontamination process generally consisted of silica (SiO{sub 2}), calcium oxide (CaO) and aluminum oxide (Al{sub 2}O{sub 3}). Also, Calcium fluoride (CaF{sub 2} ), nickel oxide (NiO), and ferric oxide (Fe{sub 2}O{sub 3}) were added to provide an increase in the slag fluidity and oxidative potential. Cerium was used as a surrogate for the uranium because the thermochemical and physical properties of cerium are very similar to those of uranium. Cerium was removed from the ingot phase to slag phase by up to 99% in this study. The absorption ratio of cerium was increased with an increase of the amount of the slag former. And the maximum removal of cerium occurred when the basicity index of the slag former was 0.82. The natural uranium (UO{sub 2}) was partitioned from the ingot phase to the slag phase by up to 95%. The absorption of the natural uranium was considerably dependent on the basicity index of the slag former and the composition of the slag former. The optimum condition for the removal of the uranium was about 1.5 for the basicity index and 15 wt% of the slag former. According to the increase of the amount of slag former, the absorption of uranium oxide in the slag phase was linearly increased due to an increase of its capacity to capture uranium oxide within the slag phase. Through experiments with various slag formers, we verified that the slag formers containing calcium fluoride (CaF{sub 2}) and a high amount of silica were more effective for a melt decontamination of stainless steel wastes contaminated with uranium. During the melting tests with stainless steel wastes from the uranium conversion plant(UCP ) in KAERI, we found that the results of the uranium decontamination were very similar to those of the uranium oxide from the melting of stimulated metal wastes. (authors)

Uranium storage practices at US Department of Energy (DOE) facilities have evolved over a period spanning five decades of programmatic work in support of the nuclear deterrent mission. During this period, the Y-12 Plant in Oak Ridge, Tennessee has served as the principal enriched uranium facility for fabrication, chemical processing, metallurgical processing and storage. Recent curtailment of new nuclear weapons production and stockpile reduction has created significant amounts of enriched uranium available as a strategic resource which must be properly and safely stored. This standard specifies criteria associated with the safe storage of enriched uranium at the Y-12 Plant. Because programmatic needs, compliance regulations and desirable materials of construction change with time, it is recommended that these standards be reviewed and amended periodically to ensure that they continue to serve their intended purpose.

As a result of the end of the Cold War, the United States nuclear weapon stockpile is being reduced from approximately 20,000 warheads to fewer than 10,000 by the end of the century. The Oak Ridge Y-12 Plant is the Department of Energy (DOE) site charged with the responsibility of providing safe, secure storage for the uranium recovered from these weapons. In addition to weapons material, Y-12 has traditionally processed and stored uranium from nonweapon programs and presumably will continue to do so. The purpose of this document is to evaluate the suitability of plastics for use in the containment of uranium.

The uranium enrichment pilot plant at PNC Ningyo-Toge Works, Japan, started operation in August 1979. Since then, inspection activities by the government of Japan and the International Atomic Energy Agency (IAEA) have been carried out. A basic measure of safeguards is evaluation of material unaccounted for (MUF) by closing the material balance. As the plant now produces uranium of <5% enrichment, a material balance is closed only once a year. Until now, eight physical inventories have been taken. This paper describes the operator's procedures for material accountability and the values of MUF reported to the government of Japan and the IAEA.

A chemical decontamination process using nitric acid solution was selected as in-situ technology for recycle or release with authorization of a large amount of metallic waste including process system components such as tanks, piping, etc., which is generated by dismantling a retired uranium conversion plant at Korea Atomic Energy Research Institute (KAERI). The applicability of nitric acid solution for surface decontamination of metallic wastes contaminated with uranium compounds was evaluated through the basic research on the dissolution of UO2 and ammonium uranyl carbonate (AUC) powder. Decontamination performance was verified by using the specimens contaminated with such uranium compounds as UO2 and AUC taken from the uranium conversion plant. Dissolution rate of UO2 powder is notably enhanced by the addition of H2O2 as an oxidant even in the condition of a low concentration of nitric acid and low temperature compared with those in a nitric acid solution without H2O2. AUC powders dissolve easily in nitric acid solutions until the solution pH attains about 2.5 {approx} 3. Above that solution pH, however, the uranium concentration in the solution is lowered drastically by precipitation as a form of U3(NH3)4O9 . 5H2O. Decontamination performance tests for the specimens contaminated with UO2 and AUC were quite successful with the application of decontamination conditions obtained through the basic studies on the dissolution of UO2 and AUC powders.

During the next two decades the transuranic wastes, now stored in the burial trenches and storage facilities at the Hanford Site, are to be retrieved, processed at the Waste Receiving and Processing Facility, and shipped to the Waste Isolation Pilot Plant near Carlsbad, New Mexico for final disposal. Over 7% of the transuranic waste to be retrieved for shipment to the Waste Isolation Pilot Plant has been generated at the Plutonium-Uranium Extraction (PUREX) Plant. The purpose of this report is to characterize the radioactive solid wastes generated by PUREX using process knowledge, existing records, and oral history interviews. The PUREX Plant is currently operated by the Westinghouse Hanford Company for the US Department of Energy and is now in standby status while being prepared for permanent shutdown. The PUREX Plant is a collection of facilities that has been used primarily to separate plutonium for nuclear weapons from spent fuel that had been irradiated in the Hanford Site`s defense reactors. Originally designed to reprocess aluminum-clad uranium fuel, the plant was modified to reprocess zirconium alloy clad fuel elements from the Hanford Site`s N Reactor. PUREX has provided plutonium for research reactor development, safety programs, and defense. In addition, the PUREX was used to recover slightly enriched uranium for recycling into fuel for use in reactors that generate electricity and plutonium. Section 2.0 provides further details of the PUREX`s physical plant and its operations. The PUREX Plant functions that generate solid waste are as follows: processing operations, laboratory analyses and supporting activities. The types and estimated quantities of waste resulting from these activities are discussed in detail.

One of the more important legacies left with the Department of Energy (DOE) after the privatization of the United States Enrichment Corporation is the large inventory of depleted uranium hexafluoride (DUF6). The DOE Office of Nuclear Energy, Science and Technology (NE) is responsible for the long-term management of some 700,000 metric tons of DUF6 stored at the sites of the two gaseous diffusion plants located at Paducah, Kentucky and Portsmouth, Ohio, and at the East Tennessee Technology Park in Oak Ridge, Tennessee. The DUF6 management program resides in NE's Office of Depleted Uranium Hexafluoride Management. The current DUF6 program has largely focused on the ongoing maintenance of the cylinders containing DUF6. However, the long-term management and eventual disposition of DUF6 is the subject of a Programmatic Environmental Impact Statement (PEIS) and Public Law 105-204. The first step for future use or disposition is to convert the material, which requires construction and long-term operation of one or more conversion plants. To help inform the DUF6 program's planning activities, it was necessary to perform design and cost studies of likely DUF6 conversion plants at the preconceptual level, beyond the PEIS considerations but not as detailed as required for conceptual designs of actual plants. This report contains the final results from such a preconceptual design study project. In this fast track, three month effort, Lawrence Livermore National Laboratory and Bechtel National Incorporated developed and evaluated seven different preconceptual design cases for a single plant. The preconceptual design, schedules, costs, and issues associated with specific DUF6 conversion approaches, operating periods, and ownership options were evaluated based on criteria established by DOE. The single-plant conversion options studied were similar to the dry-conversion process alternatives from the PEIS. For each of the seven cases considered, this report contains information on the conversion process, preconceptual plant description, rough capital and operating costs, and preliminary project schedule.

The U.S. Department of Energy developed this statement to evaluate the environmental impacts of resumption of operations of the PUREX/Uranium Oxide facilities at the Hanford Site to produce plutonium and other special nuclear materials for national defense needs.

Biogeochemistry of uranium in wetlands plays important roles in U immobilization in storage ponds of U mining and processing facilities but has not been well understood. The objective of this work was to study molecular mechanisms responsible for high U retention by Savannah River Site (SRS) wetland sediments under varying redox and acidic (pH = 2.6–5.8) conditions using U L?-edge X-ray absorption spectroscopy. Uranium in the SRS wetland sediments existed primarily as U(VI) bonded as a bidentate to carboxylic sites (U–C bond distance at ?2.88 Ĺ), rather than phenolic or other sites of natural organic matter (NOM). In microcosms simulatingmore »the SRS wetland processes, U immobilization on roots was 2 orders of magnitude higher than on the adjacent brown or more distant white sands in which U was U(VI). Uranium on the roots were both U(IV) and U(VI), which were bonded as a bidentate to carbon, but the U(VI) may also form a U phosphate mineral. After 140 days of air exposure, all U(IV) was reoxidized to U(VI) but remained as a bidentate bonding to carbon. This study demonstrated NOM and plant roots can highly immobilize U(VI) in the SRS acidic sediments, which has significant implication for the long-term stewardship of U-contaminated wetlands.« less

In the U-Pu co-denitration facility at Rokkasho Reprocessing Plant (RRP), Active Test which composes of 5 steps was performed by using uranium-plutonium nitrate solution that was extracted from spent fuels. During Active Test, two kinds of tests were performed in parallel. One was denitration performance test in denitration ovens, and expected results were successfully obtained. The other was validation and calibration of non-destructive assay (NDA) systems, and expected performances were obtained and their effectiveness as material accountancy and safeguards system was validated. (authors)

This study evaluated the abilities of various plant species to act as bio-monitors for environmental uranium (U) contamination. Vegetation and soil samples were collected from a U processing facility. The water-way fed from facility storm and processing effluents was the focal sample site as it represented a primary U transport mechanism. Soils and sediments from areas exposed to contamination possessed U concentrations that averaged 630 mg U kg{sup -1}. Aquatic mosses proved to be exceptional accumulators of U with dry weight (dw) concentrations measuring as high as 12500 mg U kg{sup -1} (approximately 1% of the dw mass was attributable to U). The macrophytes (Phragmites communis, Scripus fontinalis and Sagittaria latifolia) were also effective accumulators of U. In general, plant roots possessed higher concentrations of U than associated upper portions of plants. For terrestrial plants, the roots of Impatiens capensis had the highest observed levels of U accumulation (1030 mg kg{sup -1}), followed by the roots of Cyperus esculentus and Solidago speciosa. The concentration ratio (CR) characterized dry weight (dw) vegetative U levels relative to that in associated dw soil. The plant species that accumulated U at levels in excess of that found in the soil were: P. communis root (CR, 17.4), I. capensis root (CR, 3.1) and S. fontinalis whole plant (CR, 1.4). Seven of the highest ten CR values were found in the roots. Correlations with concentrations of other metals with U were performed, which revealed that U concentrations in the plant were strongly correlated with nickel (Ni) concentrations (correlation: 0.992; r-squared: 0.984). Uranium in plant tissue was also strongly correlated with strontium (Sr) (correlation: 0.948; r-squared: 0.899). Strontium is chemically and physically similar to calcium (Ca) and magnesium (Mg), which were also positively-correlated with U. The correlation with U and these plant nutrient minerals, including iron (Fe), suggests that active uptake mechanisms may influence plant U accumulation.

In support of the Gaseous Diffusion Plant Safety Analysis Report Upgrade program (GDP SARUP), a natural phenomena hazards evaluation was performed for the main process equipment and piping in the uranium enrichment buildings at Paducah and Portsmouth gaseous diffusion plants. In order to reduce the cost of rigorous analyses, the evaluation methodology utilized a graded approach based on an experience data base collected by SQUG/EPRI that contains information on the performance of industrial equipment and piping during past earthquakes. This method consisted of a screening walkthrough of the facility in combination with the use of engineering judgment and simple calculations. By using these screenings combined with evaluations that contain decreasing conservatism, reductions in the time and cost of the analyses were significant. A team of experienced seismic engineers who were trained in the use of the DOE SQUG/EPRI Walkdown Screening Material was essential to the success of this natural phenomena hazards evaluation.

Safety and cost information is developed for the conceptual decommissioning of a commercial uranium hexafluoride conversion (UF{sub 6}) plant. Two basic decommissioning alternatives are studied to obtain comparisons between cost and safety impacts: DECON, and passive SAFSTOR. A third alternative, DECON of the plant and equipment with stabilization and long-term care of lagoon wastes. is also examined. DECON includes the immediate removal (following plant shutdown) of all radioactivity in excess of unrestricted release levels, with subsequent release of the site for public use. Passive SAFSTOR requires decontamination, preparation, maintenance, and surveillance for a period of time after shutdown, followed by deferred decontamination and unrestricted release. DECON with stabilization and long-term care of lagoon wastes (process wastes generated at the reference plant and stored onsite during plant operation} is also considered as a decommissioning method, although its acceptability has not yet been determined by the NRC. The decommissioning methods assumed for use in each decommissioning alternative are based on state-of-the-art technology. The elapsed time following plant shutdown required to perform the decommissioning work in each alternative is estimated to be: for DECON, 8 months; for passive SAFSTOR, 3 months to prepare the plant for safe storage and 8 months to accomplish deferred decontamination. Planning and preparation for decommissioning prior to plant shutdown is estimated to require about 6 months for either DECON or passive SAFSTOR. Planning and preparation prior to starting deferred decontamination is estimated to require an additional 6 months. OECON with lagoon waste stabilization is estimated to take 6 months for planning and about 8 months to perform the decommissioning work. Decommissioning cost, in 1981 dollars, is estimated to be $5.91 million for OECON. For passive SAFSTOR, preparing the facility for safe storage is estimated to cost $0.88 million, the annual maintenance and surveillance cost is estimated to be about $0.095 million, and deferred decontamination is estimated to cost about $6.50 million. Therefore, passive SAFSTOR for 10 years is estimated to cost $8.33 million in nondiscounted 1981 dollars. DECON with lagoon waste stabilization is estimated to cost about $4.59 million, with an annual cost of $0.011 million for long-term care. All of these estimates include a 25% contingency. Waste management costs for DECON, including the net cost of disposal of the solvent extraction lagoon wastes by shipping those wastes to a uranium mill for recovery of residual uranium, comprise about 38% of the total decommissioning cost. Disposal of lagoon waste at a commercial low-level waste burial ground is estimated to add $10.01 million to decommissioning costs. Safety analyses indicate that radiological and nonradiological safety impacts from decommissioning activities should be small. The 50-year committed dose equivalent to members of the public from airborne releases during normal decommissioning activities is estimated to 'Je about 4.0 man-rem. Radiation doses to the public from accidents are found to be very low for all phases of decommissioning. Occupational radiation doses from normal decommissioning operations (excluding transport operations) are estimated to be about 79 man-rem for DECON and about 80 man-rem for passive SAFSTOR with 10 years of safe storage. Doses from DECON with lagoon waste stabilization are about the same as for DECON except there is less dose resulting from transportation of radioactive waste. The number of fatalities and serious lost-time injuries not related to radiation is found to be very small for all decommissioning alternatives. Comparison of the cost estimates shows that DECON with lagoon waste stabilization is the least expensive method. However, this alternative does not allow unrestricted release of the site. The cumulative cost of maintenance and surveillance and the higher cost of deferred decontamination makes passive SAFSTOR more expensive than DECON. Seve

This document is an assessment of the technical safety issues pertaining to the storage of EU at the Oak Ridge Y-12 Plant. The purpose of the assessment is to serve as the basis for defining the technical standards for storage of EU at Y-12. A formal assessment of the Y-12 materials acceptance criteria for EU is currently being conducted by a task force cochaired by B. G. Eddy of DOE Oak Ridge Operations and S. 0. Cox of Y-12 Defense Programs. The mission of this technical assessment for storage is obviously dependent on results of the acceptance assessment. Clearly, the two efforts require coordination to avoid inconsistencies. In addition, both these Assessments must be consistent with the Environmental Assessment for EU storage at Y-12.1 Both the Storage Assessment and the Criteria for Acceptance must take cognizance of the fact that a portion of the EU to be submitted for storage in the future is expected to be derived from foreign sources and to include previously irradiated uranium containing significant levels of transuranics, radioactive daughter products, and unstable uranium isotopes that do not occur in the EU stream of the DOE weapons complex. National security considerations may dictate that these materials be accepted despite the fact that they fail to conform to the Acceptance Criteria. This document will attempt to address the complexities inherent in this situation.

Characterization of the level and extent of groundwater contamination in the vicinity of Title I mill sites began during the surface remedial action stage (Phase 1) of the Uranium Mill Tailings Remedial Action (UMTRA) Project. Some of the contamination in the aquifer(s) at the abandoned sites is attributable to milling activities during the years the mills were in operation. To begin implementation of Phase 11 groundwater remediation, the US Department of Energy (DOE) requested that (1) the Technical Assistance Contractor (TAC) conduct a study to provide for the design of a mobile water treatment plant to treat groundwater extracted during site characterization studies at completed Phase I UMTRA sites, and (2) the results of the TAC investigations be documented in a special study report. This special study develops the design criteria for a water treatment plant that can be readily transported from one UMTRA site to another and operated as a complete treatment system. The 1991 study provides the basis for selecting a mobile water treatment system to meet the operating requirements recommended in this special study. The scope of work includes the following: Determining contaminants, flows, and loadings. Setting effluent quality criteria. Sizing water treatment unit(s). Evaluating non-monetary aspects of alternate treatment processes. Comparing costs of alternate treatment processes. Recommending the mobile water treatment plant design criteria.

Recent advances in radio frequency identification devices (RFIDs) have engendered a growing interest among international safeguards experts. Potentially, RFIDs could reduce inspection work, viz. the number of inspections, number of samples, and duration of the visits, and thus improve the efficiency and effectiveness of international safeguards. This study systematically examined the applications of RFIDs for IAEA safeguards at large gas-centrifuge enrichment plants (GCEPs). These analyses are expected to help identify the requirements and desirable properties for RFIDs, to provide insights into which vulnerabilities matter most, and help formulate the required assurance tests. This work, specifically assesses the application of RFIDs for the ''Option 4'' safeguards approach, proposed by Bruce Moran, U. S. Nuclear Regulatory Commission (NRC), for large gas-centrifuge uranium-enrichment plants. The features of ''Option 4'' safeguards include placing RFIDs on all feed, product and tails (F/P/T) cylinders, along with WID readers in all FP/T stations and accountability scales. Other features of Moran's ''Option 4'' are Mailbox declarations, monitoring of load-cell-based weighing systems at the F/P/T stations and accountability scales, and continuous enrichment monitors. Relevant diversion paths were explored to evaluate how RFIDs improve the efficiency and effectiveness of safeguards. Additionally, the analysis addresses the use of RFIDs in conjunction with video monitoring and neutron detectors in a perimeter-monitoring approach to show that RFIDs can help to detect unidentified cylinders.

Over a two and one-half year period beginning in 1981, representatives of six countries (United States, United Kingdom, Federal Republic of Germany, Australia, The Netherlands, and Japan) and the inspectorate organizations of the International Atomic Energy Agency and EURATOM developed and agreed to a technically sound approach for verifying the absence of highly enriched uranium (HEU) in gas centrifuge enrichment plants. This effort, known as the Hexapartite Safeguards Project (HSP), led to the first international concensus on techniques and requirements for effective verification of the absence of weapons-grade nuclear materials production. Since that agreement, research and development has continued on the radiation detection technology-based technique that technically confirms the HSP goal is achievable. However, the realities of achieving the HSP goal of effective technical verification have not yet been fully attained. Issues such as design and operating conditions unique to each gas centrifuge plant, concern about the potential for sensitive technology disclosures, and on-site support requirements have hindered full implementation and operator support of the HSP agreement. In future arms control treaties that may limit or monitor fissile material production, the negotiators must recognize and account for the realities and practicalities in verifying the absence of HEU production. This paper will describe the experiences and realities of trying to achieve the goal of developing and implementing an effective approach for verifying the absence of HEU production. 3 figs.

Decommissioning of a nuclear facility comprises not only actual dismantling but also, above all, management of the resulting residual materials and waste. Siemens Decommissioning Projects (DP) in Hanau has been involved in this task since 1995 when the decision was taken to decommission and dismantle the Hanau Fuel Fabrication Plant. Due to the decommissioning, large amounts of contaminated steel scrap have to be managed. The contamination of this metal scrap can be found almost exclusively in the form of surface contamination. Various decontamination technologies are involved, as there are blasting and wiping. Often these methods are not sufficient to meet the free release limits. In these cases, SIEMENS has decided to melt the scrap at Siempelkamp's melting plant. The plant is licensed according to the German Radiation Protection Ordinance Section 7 (issue of 20.07.2001). The furnace is a medium frequency induction type with a load capacity of 3.2 t and a throughput of 2 t/h for steel melting. For safety reasons, the furnace is widely operated by remote handling. A highly efficient filter system of cyclone, bag filter and HEPA-filter in two lines retains the dust and aerosol activity from the off-gas system. The slag is solidified at the surface of the melt and gripped before pouring the liquid iron into a chill. Since 1989, in total 15,000 t have been molten in the plant, 2,000 t of them having been contaminated steel scrap from the decommissioning of fuel fabrication plants. Decontamination factors could be achieved between 80 and 100 by the high affinity of the uranium to the slag former. The activity is transferred to the slag up to nearly 100 %. Samples taken from metal, slag and dust are analyzed by gamma measurements of the 186 keV line of U235 and the 1001 keV line of Pa234m for U238. All produced ingots showed a remaining activity less than 1 Bq/g and could be released for industrial reuse.

For the past two years, the United States National Nuclear Security Administration, Office of International Regimes and Agreements (NA-243), has sponsored the Safeguards-by-Design Project, through which it is hoped new nuclear facilities will be designed and constructed worldwide more amenable to nuclear safeguards. In the course of this project it was recognized that commercial designer/builders of nuclear facilities are not always aware of, or understand, the relevant domestic and international safeguards requirements, especially the latter as implemented by the International Atomic Energy Agency (IAEA). To help commercial designer/builders better understand these requirements, a report was prepared by the Safeguards-by-Design Project Team that articulated and interpreted the international nuclear safeguards requirements for the initial case of uranium enrichment plants. The following paper summarizes the subject report, the specific requirements, where they originate, and the implications for design and construction. It also briefly summarizes the established best design and operating practices that designer/builder/operators have implemented for currently meeting these requirements. In preparing the subject report, it is recognized that the best practices are continually evolving as the designer/builder/operators and IAEA consider even more effective and efficient means for meeting the safeguards requirements and objectives.

The objective of safeguards is the timely detection of the diversion of a significant quantity of nuclear materials, and safeguarding uranium enrichment plants is especially important in preventing the spread of nuclear weapons. The IAEA’s proposed Unattended Cylinder Verification Station (UCVS) for UF6 cylinder verification would combine the operator’s accountancy scale with a nondestructive assay system such as the Passive Neutron Enrichment Meter (PNEM) and cylinder identification and surveillance systems. In this project, we built a laboratory-scale UCVS and demonstrated its capabilities using mock UF6 cylinders. We developed a signal processing algorithm to automate the data collection and processing from four continuous, unattended sensors. The laboratory demonstration of the system showed that the software could successfully identify cylinders, snip sensor data at the appropriate points in time, determine the relevant characteristics of the cylinder contents, check for consistency among sensors, and output the cylinder data to a file. This paper describes the equipment, algorithm and software development, laboratory demonstration, and recommendations for a full-scale UCVS.

Enhancement of the nuclear materials protection, control, and accounting of (MPC&A) at the Elektrostal Machine-Building Plant (ELEMASH) has proceeded in two phases. Initially, Elektrostal served as the model facility at which to test US/Russian collaboration and to demonstrate MPC&A technologies available for safeguards enhancements at Russian facilities. This phase addressed material control and accounting (MC&A) in the low-enriched uranium (LEU) fuel-fabrication processes and the physical protection (PP) of part of the (higher-enrichment) breeder-fuel process. The second phase, identified later in the broader US/Russian agreement for expanded MPC&A cooperation. includes implementation of appropriate MC&A and PP systems in the breeder-fuel fabrication processes. Within the past year, an automated physical protection system has been installed and demonstrated in building 274, and an automated MC&A system has been designed and is being installed and will be tested in the LEU process. Attention has now turned to assuring longterm sustainability for the first phase and beginning MPC&A upgrades for the second phase. Sustainability measures establish the infrastructure for operation, maintenance, and repair of the installed systems-with US support for the lifetime of the US/Russian Agreement, but evolving toward full Russian operation of the system over the long term. For phase 2, which will address higher enrichments, projects have been identified to characterize the facilities, design MPC&A systems, procure appropriate equipment, and install and test final systems. One goal in phase 2 will be to build on initial work to create shared, plant-wide MPC&A assets for operation, maintenance, and evaluation of all safeguards systems.

Uranium enrichment in the United States has utilized a diffusion process to preferentially enrich the U-235 isotope in the uranium product. The U-AVLIS process is based on electrostatic extraction of photoionized U-235 atoms from an atomic vapor stream created by electron-beam vaporization of uranium metal alloy. The U-235 atoms are ionized when precisely tuned laser light -- of appropriate power, spectral, and temporal characteristics -- illuminates the uranium vapor and selectively photoionizes the U-235 isotope. A programmatic document for use in screening DOE site to locate a U-AVLIS production plant was developed and implemented in two parts. The first part consisted of a series of screening analyses, based on exclusionary and other criteria, that identified a reasonable number of candidate sites. These sites were subjected to a more rigorous and detailed comparative analysis for the purpose of developing a short list of reasonable alternative sites for later environmental examination. This environmental site description (ESD) provides a detailed description of the PGDP site and vicinity suitable for use in an environmental impact statement (EIS). The report is based on existing literature, data collected at the site, and information collected by Argonne National Laboratory (ANL) staff during a site visit. 65 refs., 15 tabs.

International Atomic Energy Agency (IAEA) inspectors currently perform periodic inspections at uranium enrichment plants to verify UF6 cylinder enrichment declarations. Measurements are typically performed with handheld high-resolution sensors on a sampling of cylinders taken to be representative of the facility’s entire cylinder inventory. These measurements are time-consuming, expensive, and assay only a small fraction of the total cylinder volume. An automated nondestructive assay system capable of providing enrichment measurements over the full volume of the cylinder could improve upon current verification practices in terms of manpower and assay accuracy. Such a station would use sensors that can be operated in an unattended mode at an industrial facility: medium-resolution scintillators for gamma-ray spectroscopy (e.g., NaI(Tl)) and moderated He-3 neutron detectors. This sensor combination allows the exploitation of additional, more-penetrating signatures beyond the traditional 185-keV emission from U-235: neutrons produced from F-19(?,n) reactions (spawned primarily from U 234 alpha emission) and high-energy gamma rays (extending up to 8 MeV) induced by neutrons interacting in the steel cylinder. This paper describes a study of these non-traditional signatures for the purposes of cylinder enrichment verification. The signatures and the radiation sensors designed to collect them are described, as are proof-of-principle cylinder measurements and analyses. Key sources of systematic uncertainty in the non-traditional signatures are discussed, and the potential benefits of utilizing these non-traditional signatures, in concert with an automated form of the traditional 185-keV-based assay, are discussed.

Uranium enrichment in the United States has utilized a diffusion process to preferentially enrich the U-235 isotope in the uranium product. In the 1970s, the US Department of Energy (DOE) began investigating more efficient and cost-effective enrichment technologies. In January 1990, the Secretary of Energy approved a plan for the demonstration and deployment of the Uranium Atomic Vapor Laser isotope Separation (U-AVLIS) technology with the near-term goal to provide the necessary information to make a deployment decision by November 1992. Initial facility operation is anticipated for 1999. A programmatic document for use in screening DOE sites to locate a U-AVLIS production plant was developed and implemented in two parts. The first part consisted of a series of screening analyses, based on exclusionary and other criteria, that identified a reasonable number of candidate sites. The final evaluation, which included sensitivity studies, identified the Oak Ridge Gaseous Diffusion Plant (ORGDP) site, the Paducah Gaseous Diffusion Plant (PGDP) site, and the Portsmouth Gaseous Diffusion Plant (PORTS) site as having significant advantages over the other sites considered. This environmental site description (ESD) provides a detailed description of the PORTS site and vicinity suitable for use in an environmental impact statement (EIS). This report is based on existing literature, data collected at the site, and information collected by Argonne National Laboratory (ANL) staff during site visits. The organization of the ESD is as follows. Topics addressed in Sec. 2 include a general site description and the disciplines of geology, water resources, biotic resources, air resources, noise, cultural resources, land use. Socioeconomics, and waste management. Identification of any additional data that would be required for an EIS is presented in Sec. 3.

The sintering behavior of uranium and uranium-zirconium alloys in the alpha phase were characterized in this research. Metal uranium powder was produced from pieces of depleted uranium metal acquired from the Y-12 plant via hydriding...

The sintering behavior of uranium and uranium-zirconium alloys in the alpha phase were characterized in this research. Metal uranium powder was produced from pieces of depleted uranium metal acquired from the Y-12 plant via hydriding...

Some of nuclear facilities that would no longer be required have been decommissioned in JAEA (Japan Atomic Energy Agency). A lot of nuclear facilities have to be decommissioned in JAEA in near future. To implement decommissioning of nuclear facilities, it was important to make a rational decommissioning plan. Therefore, project management data evaluation system for dismantling activities (PRODIA code) has been developed, and will be useful for making a detailed decommissioning plan for an object facility. Dismantling of dry conversion facility in the uranium refining and conversion plant (URCP) at Ningyo-toge began in 2008. During dismantling activities, project management data such as manpower and amount of waste generation have been collected. Such collected project management data has been evaluated and used to establish a calculation formula to calculate manpower for dismantling equipment of chemical process and calculate manpower for using a green house (GH) which was a temporary structure for preventing the spread of contaminants during dismantling. In the calculation formula to calculate project management data related to dismantling of equipment, the relation of dismantling manpower to each piece of equipment was evaluated. Furthermore, the relation of dismantling manpower to each chemical process was evaluated. The results showed promise for evaluating dismantling manpower with respect to each chemical process. In the calculation formula to calculate project management data related to use of the GH, relations of GH installation manpower and removal manpower to GH footprint were evaluated. Furthermore, the calculation formula for secondary waste generation was established. In this study, project management data related to dismantling of equipment and use of the GH were evaluated and analyzed. The project management data, manpower for dismantling of equipment, manpower for installation and removal of GH, and secondary waste generation from GH were considered. Establishment of the calculation formula for dismantling of each kind of equipment makes it possible to evaluate manpower for dismantling the whole facility. However, it is not easy to prepare calculation formula for all kinds of equipment that exist in the facility. Therefore, a simpler evaluation method was considered to calculate manpower based on facility characteristics. The results showed promise for evaluating dismantling manpower with respect to each chemical process. For dismantling of contaminated equipment, a GH has been used for protection of the spread of contamination. The use of a GH increases manpower for installation and removal of GH etc. Moreover, structural materials of the GH such as plastic sheets, adhesive tape become a burnable secondary waste. To create an effective dismantling plan, it is necessary to carefully consider use of a GH preliminarily. Thus, an evaluation method of project management data such as manpower and secondary waste generation was considered. The results showed promise for evaluating project management data of GH by using established calculation formula. (authors)

Approved industry-standard cylinders are used globally to handle and store uranium hexafluoride (UF{sub 6}) feed, product, tails, and samples at uranium enrichment plants. The International Atomic Energy Agency (IAEA) relies on time-consuming physical inspections to verify operator declarations and detect possible diversion of UF{sub 6}. Development of a reliable, automated, and tamper-resistant system for near real-time tracking and monitoring UF{sub 6} cylinders (as they move within an enrichment facility) would greatly improve the inspector function. This type of system can reduce the risk of false or misreported cylinder tare weights, diversion of nuclear material, concealment of excess production, utilization of undeclared cylinders, and misrepresentation of the cylinders contents. This paper will describe a proof-of-concept approach that was designed to evaluate the feasibility of using radio frequency (RF)-based technologies to track individual UF{sub 6} cylinders throughout a portion of their life cycle, and thus demonstrate the potential for improved domestic accountability of materials, and a more effective and efficient method for application of site-level IAEA safeguards. The evaluation system incorporates RF-based identification devices (RFID) which provide a foundation for establishing a reliable, automated, and near real-time tracking system that can be set up to utilize site-specific, rules-based detection algorithms. This paper will report results from a proof-of-concept demonstration at a real enrichment facility that is specifically designed to evaluate both the feasibility of using RF to track cylinders and the durability of the RF equipment to survive the rigors of operational processing and handling. The paper also discusses methods for securely attaching RF devices and describes how the technology can effectively be layered with other safeguard systems and approaches to build a robust system for detecting cylinder diversion. Additionally, concepts for off-site tracking of cylinders are described.

The Oak Ridge Y-12 Plant is situated adjacent to the Oak Ridge city limits and is operated by the United States Department of Energy (DOE). The Y-12 Plant is located on 4,860 acres, which is collectively referred to as the Y-12 Plant site. Among the missions for which the facility is in existence are producing nuclear weapons components, supporting weapon design laboratories, and processing special nuclear materials (SNM). The Y-12 Plant is under the regulatory guidance of DOE Order 5400.5 and has complied with the technical requirements governing SNM since its issue. However, an in-depth review with appropriate documentation had not been performed, prior to the effect presented herein, to substantiate this claim. As a result of the solid waste issue, it was determined that other types of waste should be formally reviewed for content with respect to SNM. Therefore, a project was formed to investigate the conveyance of SNM through the sanitary sewer system. It is emphasized that this project addresses only effluent from the sanitary sewer system and not the storm sewer system. The project reviewed sanitary sewer data both for the Y-12 Plant and the Y-12 Plant site.

In January 1990, the Secretary of Energy approved a plan for the demonstration and deployment of the Uranium Atomic Vapor Laser Isotope Separation (U-AVLIS) technology, with the near-term goal to provide the necessary information to make a deployment decision by November 1992. The U-AVLIS process is based on electrostatic extraction of photoionized U-235 atoms from an atomic vapor stream created by electron-beam vaporization of uranium metal alloy. A programmatic document for use in screening DOE sites to locate the U-AVLIS production plant was developed and implemented in two parts (Wolsko et al. 1991). The first part consisted of a series of screening analyses, based on exclusionary and other criteria, that identified a reasonable number of candidate sites. These sites were then subjected to a more rigorous and detailed comparative analysis for the purpose of developing a short list of reasonable alternative sites for later environmental examination. This environmental site description (ESD) provides a detailed description of the ORGDP site and vicinity suitable for use in an environmental impact statement (EIS). The report is based on existing literature, data collected at the site, and information collected by Argonne National Laboratory (ANL) staff during a site visit. The organization of the ESD is as follows. Topics addressed in Sec. 2 include a general site description and the disciplines of geology, water resources, biotic resources, air resources, noise, cultural resources, land use, socioeconomics, and waste management. Identification of any additional data that would be required for an EIS is presented in Sec. 3. Following the site description and additional data requirements, Sec. 4 provides a short, qualitative assessment of potential environmental issues. 37 refs., 20 figs., 18 tabs.

The Paducah Gaseous Diffusion Plant (PGDP) is a uranium enrichment facility owned by the US Department of Energy (DOE). A residual of the uranium enrichment process is depleted uranium hexafluoride (UF6). Depleted UF6, a solid at ambient temperature, is stored in 32,200 steel cylinders that hold a maximum of 14 tons each. Storage conditions are suboptimal and have resulted in accelerated corrosion of cylinders, increasing the potential for a release of hazardous substances. Consequently, the DOE is proposing refurbishment of certain existing yards and construction of a new storage yard. This environmental assessment (EA) evaluates the impacts of the proposed action and no action and considers alternate sites for the proposed new storage yard. The proposed action includes (1) renovating five existing cylinder yards; (2) constructing a new UF6 storage yard; handling and onsite transport of cylinders among existing yards to accommodate construction; and (4) after refurbishment and construction, restacking of cylinders to meet spacing and inspection requirements. Based on the results of the analysis reported in the EA, DOE has determined that the proposed action is not a major Federal action that would significantly affect the quality of the human environment within the context of the National Environmental Policy Act of 1969. Therefore, DOE is issuing a Finding of No Significant Impact. Additionally, it is reported in this EA that the loss of less than one acre of wetlands at the proposed project site would not be a significant adverse impact.

The US and UK have been separately working on the development of a NDA instrument to determine the enrichment of gaseous UF/sub 6/ at low pressures in cascade header pipework in line with the conclusions of the Hexapartite Safeguards Project viz. the instrument is capable of making a ''go/no go'' decision of whether the enrichment is less than/greater than 20%. Recently, there has been a series of very useful technical exchanges of ideas and information between the two countries. This has led to a technical formulation for such an instrumentation based on ..gamma..-ray spectrometry which, although plant-specific in certain features, nevertheless is based on the same physical principles. Experimental results from commercially operating enrichment plants are very encouraging and indicate that a complete measurement including set up time on the pipe should be attainable in about 30 minutes when measuring pipes of diameter around 110 mm. 5 refs., 4 figs.

One of the salient features of the transition plan that the United States Department of Energy/National Nuclear Security Administration (DOE/NNSA) is presently implementing in the Former Soviet Union countries is the availability of uranium certified reference materials for calibration of nondestructive assay (NDA) measurement equipment. To address this challenge, DOE/NNSA and U.S. national laboratories have focused their cooperative efforts on establishing a reliable source for manufacturing, certifying, and supplying of such standards. The Ulba Metallurgical Plant (UMP), Kazakhstan, which processes large quantities of low-enriched uranium to produce ceramic fuel pellets for nuclear-powered reactors, is well situated to become a key supplier of low-enriched uranium certified reference materials for the country and Central Asia region. We have recently completed Phase I of a feasibility study to establish at UMP capabilities of manufacturing these standards. In this paper we will discuss details of a proposed methodology for uranium down-blending, material selection and characterization, and a proposed methodology of measurement by destructive (DA) and non-destructive (NDA) analysis to form a database for material certification by the competent State authorities in the Republic of Kazakhstan. In addition, we will discuss the prospect for manufacturing of such standards at UMP.

A feasibility study has been performed to determine the potential usefulness of Compton imaging as a tool for design information verification (DIV) of uranium enrichment plants. Compton imaging is a method of gamma-ray imaging capable of imaging with a 360-degree field of view over a broad range of energies. These systems can image a room (with a time span on the order of one hour) and return a picture of the distribution and composition of radioactive material in that room. The effectiveness of Compton imaging depends on the sensitivity and resolution of the instrument as well the strength and energy of the radioactive material to be imaged. This study combined measurements and simulations to examine the specific issue of UF{sub 6} gas flow in pipes, at various enrichment levels, as well as hold-up resulting from the accumulation of enriched material in those pipes. It was found that current generation imagers could image pipes carrying UF{sub 6} in less than one hour at moderate to high enrichment. Pipes with low enriched gas would require more time. It was also found that hold-up was more amenable to this technique and could be imaged in gram quantities in a fraction of an hour. another questions arises regarding the ability to separately image two pipes spaced closely together. This depends on the capabilities of the instrument in question. These results are described in detail. In addition, suggestions are given as to how to develop Compton imaging as a tool for DIV.

An active-passive {sup 252}Cf shuffler instrument, installed and certified several years ago at Los Alamos National Laboratory's plutonium facility, has now been calibrated for different matrices to measure Waste Isolation Pilot Plant (WIPP)-destined transuranic (TRU) waste. Little or no data currently exist for these types of measurements in plant environments where sudden large changes in the neutron background radiation can significantly distort the results. Measurements and analyses of twenty-two 55-gallon drums, consisting of mixtures of varying quantities of uranium and plutonium in mostly noncombustible matrices, have been recently completed at the plutonium facility. The calibration and measurement techniques, including the method used to separate out the plutonium component, will be presented and discussed. Calculations used to adjust for differences in uranium enrichment from that of the calibration standards will be shown. Methods used to determine various sources of both random and systematic error will be indicated. Particular attention will be directed to those problems identified as arising from the plant environment. The results of studies to quantify the aforementioned distortion effects in the data will be presented. Various solution scenarios will be outlined, along with those adopted here.

Uranium hexafluoride is a dense, reactive gas used in Gaseous Diffusion Plants (GDPs) to make uranium enriched in the {sup 235}U isotope. Large quantities of UF{sub 6} exist at the GDPs in the form of in-process gas and as a solid in storage cylinders; smaller amounts exist as hot liquid during transfer operations. If liquid UF{sub 6} is released to the environment, it immediately flashes to a solid and a dense gas that reacts rapidly with water vapor in the air to form solid particles of uranyl fluoride and hydrogen fluoride gas. Preliminary analyses were done on various accidental release scenarios to determine which scenarios must be considered in the safety analyses for the GDPS. These scenarios included gas releases due to failure of process equipment and liquid/gas releases resulting from a breach of transfer piping from a cylinder. A major goal of the calculations was to estimate the response time for mitigating actions in order to limit potential off-site consequences of these postulated releases. The HGSYSTEM/UF{sub 6} code was used to assess the consequences of these release scenarios. Inputs were developed from release calculations which included two-phase, choked flow followed by expansion to atmospheric pressure. Adjustments were made to account for variable release rates and multiple release points. Superpositioning of outputs and adjustments for exposure time were required to evaluate consequences based on health effects due to exposures to uranium and HF at a specific location.

As a result of NSG restrictions, India cannot import the natural uranium required to fuel its Pressurized Heavy Water Reactors (PHWRs); consequently, it is forced to rely on the expediency of domestic uranium production. However, domestic production from mines and byproduct sources has not kept pace with demand from commercial reactors. This shortage has been officially confirmed by the Indian Planning Commission’s Mid-Term Appraisal of the country’s current Five Year Plan. The report stresses that as a result of the uranium shortage, Indian PHWR load factors have been continually decreasing. The Uranium Corporation of India Ltd (UCIL) operates a number of underground mines in the Singhbhum Shear Zone of Jharkhand, and it is all processed at a single mill in Jaduguda. UCIL is attempting to aggrandize operations by establishing new mines and mills in other states, but the requisite permit-gathering and development time will defer production until at least 2009. A significant portion of India’s uranium comes from byproduct sources, but a number of these are derived from accumulated stores that are nearing exhaustion. A current maximum estimate of indigenous uranium production is 430t/yr (230t from mines and 200t from byproduct sources); whereas, the current uranium requirement for Indian PHWRs is 455t/yr (depending on plant capacity factor). This deficit is exacerbated by the additional requirements of the Indian weapons program. Present power generation capacity of Indian nuclear plants is 4350 MWe. The power generation target set by the Indian Department of Atomic Energy (DAE) is 20,000 MWe by the year 2020. It is expected that around half of this total will be provided by PHWRs using indigenously supplied uranium with the bulk of the remainder provided by breeder reactors or pressurized water reactors using imported low-enriched uranium.

Electron beam techniques have been used to characterize uranium-contaminated soils and the Fernald Site, Ohio. Uranium particulates have been deposited on the soil through chemical spills and from the operation of an incinerator plant on the site. The major uranium phases have been identified by electron microscopy as uraninite, autunite, and uranium phosphite [U(PO{sub 3}){sub 4}]. Some of the uranium has undergone weathering resulting in the redistribution of uranium within the soil.

The results of a radiological survey conducted at the site of a former uranium recovery pilot plant operated by the Virginia-Carolina Chemical Corporation is presented. All that remains of this operation is a concrete pad situated within the boundary of a phosphate products plant now operated by Conserv, Inc., at the Nichols, Florida site. The survey included measurements designed to characterize the residual radioactivity in the vicinity of this pilot plant and to compare the quantities with federal guidelines for the release of decontaminated property for unrestricted use. The results of this survey indicate that only small quantities of radioactivity exist above normal background levels for that area. Some soil contamination was found in the vicinity of a concrete pad on which the pilot plant stood. Much of this contamination was due to /sup 226/Ra and /sup 238/U. Some beta-gamma dose rates in excess of applicable guidelines were observed in this same area. External gamma-ray exposure rates at 1 m above the ground range from 20 to 100 ..mu..R/hr. None of the direct measurements of alpha contamination were above guideline levels.

In September 1993, an implementing agreement was signed that authorized collaborative projects to enhance Russian national materials control and accounting, physical protection, and regulatory activities, with US assistance funded by the Nunn-Lugar Act. At the first US-Russian technical working group meeting in Moscow in February 1994, it was decided to identify a model facility where materials protection, control, and accounting (MPC and A) and regulatory projects could be carried out using proven technologies and approaches. The low-enriched uranium (LEU or RBMK and VVER) fuel-fabrication process at Elektrostal was selected, and collaborative work began in June 1994. Based on many factors, including initial successes at Elektrostal, the Russians expanded the cooperation by proposing five additional sites for MPC and A development: the Elektrostal medium-enriched uranium (MEU or BN) fuel-fabrication process and additional facilities at Podolsk, Dmitrovgrad, Obninsk, and Mayak. Since that time, multilaboratory teams have been formed to develop and implement MPC and A upgrades at the additional sites, and much new work is underway. This paper summarizes the current status of MPC and A enhancement projects in the LEU fuel-fabrication process and discusses the status of work that addresses similar enhancements in the MEU (BN) fuel processes at Elektrostal, under the recently expanded US-Russian MPC and A cooperation.

A computer program entitled URPE (Uranium Recovery Performance and Economics) has been developed to simulate the engineering performance and provide an economic analysis O of a plant recovering uranium from seawater. The ...

The United States Department of Energy, Oak Ridge Field Office, and Martin Marietta Energy Systems, Inc., are co-sponsoring this Second International Conference on Uranium Hexafluoride Handling. The conference is offered as a forum for the exchange of information and concepts regarding the technical and regulatory issues and the safety aspects which relate to the handling of uranium hexafluoride. Through the papers presented here, we attempt not only to share technological advances and lessons learned, but also to demonstrate that we are concerned about the health and safety of our workers and the public, and are good stewards of the environment in which we all work and live. These proceedings are a compilation of the work of many experts in that phase of world-wide industry which comprises the nuclear fuel cycle. Their experience spans the entire range over which uranium hexafluoride is involved in the fuel cycle, from the production of UF{sub 6} from the naturally-occurring oxide to its re-conversion to oxide for reactor fuels. The papers furnish insights into the chemical, physical, and nuclear properties of uranium hexafluoride as they influence its transport, storage, and the design and operation of plant-scale facilities for production, processing, and conversion to oxide. The papers demonstrate, in an industry often cited for its excellent safety record, continuing efforts to further improve safety in all areas of handling uranium hexafluoride. Selected papers were processed separately for inclusion in the Energy Science and Technology Database.

Approved industry-standard cylinders are used globally for processing, storing, and transporting uranium hexafluoride (UF{sub 6}) at uranium enrichment plants. To ensure that cylinder movements at enrichment facilities occur as declared, the International Atomic Energy Agency (IAEA) must conduct time-consuming periodic physical inspections to validate facility records, cylinder identity, and containment. By using a robust system design that includes the capability for real-time unattended monitoring (of cylinder movements), site-specific rules-based event detection algorithms, and the capability to integrate with other types of monitoring technologies, one can build a system that will improve overall inspector effectiveness. This type of monitoring system can provide timely detection of safeguard events that could be used to ensure more timely and appropriate responses by the IAEA. It also could reduce reliance on facility records and have the additional benefit of enhancing domestic safeguards at the installed facilities. This paper will discuss the installation and evaluation of a radio-frequency- (RF-) based cylinder tracking system that was installed at a United States Enrichment Corporation Centrifuge Facility. This system was installed primarily to evaluate the feasibility of using RF technology at a site and the operational durability of the components under harsh processing conditions. The installation included a basic system that is designed to support layering with other safeguard system technologies and that applies fundamental rules-based event processing methodologies. This paper will discuss the fundamental elements of the system design, the results from this site installation, and future efforts needed to make this technology ready for IAEA consideration.

In the wake of the Fukushima Daiichi nuclear power plant accident, questions are being raised about the future of the uranium market, including as regards the number of reactors expected to be built in the coming years, the amount of uranium required to meet forward demand, the adequacy of identified uranium resources to meet that demand and the ability of the sector to meet reactor requirements in a challenging investment climate. This 24th edition of the “Red Book”, a recognised world reference on uranium jointly prepared by the OECD Nuclear Energy Agency and the International Atomic Energy Agency, provides analyses and information from 42 producing and consuming countries in order to address these and other questions. It offers a comprehensive review of world uranium supply and demand as well as data on global uranium exploration, resources, production and reactor-related requirements. It also provides substantive new information on established uranium production centres around the world and in countri...

Historic data from many countries demonstrate that on average no more than 50-70% of the uranium in a deposit could be mined. An analysis of more recent data from Canada and Australia leads to a mining model with an average deposit extraction lifetime of 10+- 2 years. This simple model provides an accurate description of the extractable amount of uranium for the recent mining operations. Using this model for all larger existing and planned uranium mines up to 2030, a global uranium mining peak of at most 58 +- 4 ktons around the year 2015 is obtained. Thereafter we predict that uranium mine production will decline to at most 54 +- 5 ktons by 2025 and, with the decline steepening, to at most 41 +- 5 ktons around 2030. This amount will not be sufficient to fuel the existing and planned nuclear power plants during the next 10-20 years. In fact, we find that it will be difficult to avoid supply shortages even under a slow 1%/year worldwide nuclear energy phase-out scenario up to 2025. We thus suggest that a worldwide nuclear energy phase-out is in order. If such a slow global phase-out is not voluntarily effected, the end of the present cheap uranium supply situation will be unavoidable. The result will be that some countries will simply be unable to afford sufficient uranium fuel at that point, which implies involuntary and perhaps chaotic nuclear phase-outs in those countries involving brownouts, blackouts, and worse.

A novel process for recovering uranium from seawater is proposed and some of the critical technical parameters are evaluated. The process, in summary, consists of two different options for contacting adsorbant pellets with seawater without pumping the seawater. It is expected that this will reduce the mass handling requirements, compared to pumped seawater systems, by a factor of approximately 10/sup 5/, which should also result in a large reduction in initial capital investment. Activated carbon, possibly in combination with a small amount of dissolved titanium hydroxide, is expected to be the preferred adsorbant material instead of the commonly assumed titanium hydroxide alone. The activated carbon, after exposure to seawater, can be stripped of uranium with an appropriate eluant (probably an acid) or can be burned for its heating value (possible in a power plant) leaving the uranium further enriched in its ash. The uranium, representing about 1% of the ash, is then a rich ore and would be recovered in a conventional manner. Experimental results have indicated that activated carbon, acting alone, is not adequately effective in adsorbing the uranium from seawater. We measured partition coefficients (concentration ratios) of approximately 10/sup 3/ in seawater instead of the reported values of 10/sup 5/. However, preliminary tests carried out in fresh water show considerable promise for an extraction system that uses a combination of dissolved titanium hydroxide (in minute amounts) which forms an insoluble compound with the uranyl ion, and the insoluble compound then being sorbed out on activated carbon. Such a system showed partition coefficients in excess of 10/sup 5/ in fresh water. However, the system was not tested in seawater.

With several countries currently building nuclear power plants and planning the construction of more to meet long-term increases in electricity demand, uranium resources, production and demand remain topics of notable interest. In response to the projected growth in demand for uranium and declining inventories, the uranium industry – the first critical link in the fuel supply chain for nuclear reactors – is boosting production and developing plans for further increases in the near future. Strong market conditions will, however, be necessary to trigger the investments required to meet projected demand. The "Red Book", jointly prepared by the OECD Nuclear Energy Agency and the International Atomic Energy Agency, is a recognised world reference on uranium. It is based on information compiled in 40 countries, including those that are major producers and consumers of uranium. This 23rd edition provides a comprehensive review of world uranium supply and demand as of 1 January 2009, as well as data on global ur...

This report is the second in a series of guidelines on international safeguards requirements and practices, prepared expressly for the designers of nuclear facilities. The first document in this series is the description of generic international nuclear safeguards requirements pertaining to all types of facilities. These requirements should be understood and considered at the earliest stages of facility design as part of a new process called “Safeguards-by-Design.” This will help eliminate the costly retrofit of facilities that has occurred in the past to accommodate nuclear safeguards verification activities. The following summarizes the requirements for international nuclear safeguards implementation at enrichment plants, prepared under the Safeguards by Design project, and funded by the U.S. Department of Energy (DOE) National Nuclear Security Administration (NNSA), Office of NA-243. The purpose of this is to provide designers of nuclear facilities around the world with a simplified set of design requirements and the most common practices for meeting them. The foundation for these requirements is the international safeguards agreement between the country and the International Atomic Energy Agency (IAEA), pursuant to the Treaty on the Non-proliferation of Nuclear Weapons (NPT). Relevant safeguards requirements are also cited from the Safeguards Criteria for inspecting enrichment plants, found in the IAEA Safeguards Manual, Part SMC-8. IAEA definitions and terms are based on the IAEA Safeguards Glossary, published in 2002. The most current specification for safeguards measurement accuracy is found in the IAEA document STR-327, “International Target Values 2000 for Measurement Uncertainties in Safeguarding Nuclear Materials,” published in 2001. For this guide to be easier for the designer to use, the requirements have been restated in plainer language per expert interpretation using the source documents noted. The safeguards agreement is fundamentally a legal document. As such, it is written in a legalese that is understood by specialists in international law and treaties, but not by most outside of this field, including designers of nuclear facilities. For this reason, many of the requirements have been simplified and restated. However, in all cases, the relevant source document and passage is noted so that readers may trace the requirement to the source. This is a helpful living guide, since some of these requirements are subject to revision over time. More importantly, the practices by which the requirements are met are continuously modernized by the IAEA and nuclear facility operators to improve not only the effectiveness of international nuclear safeguards, but also the efficiency. As these improvements are made, the following guidelines should be updated and revised accordingly.

The Uranium Industry Annual 1996 (UIA 1996) provides current statistical data on the US uranium industry`s activities relating to uranium raw materials and uranium marketing. The UIA 1996 is prepared for use by the Congress, Federal and State agencies, the uranium and nuclear electric utility industries, and the public. Data on uranium raw materials activities for 1987 through 1996 including exploration activities and expenditures, EIA-estimated reserves, mine production of uranium, production of uranium concentrate, and industry employment are presented in Chapter 1. Data on uranium marketing activities for 1994 through 2006, including purchases of uranium and enrichment services, enrichment feed deliveries, uranium fuel assemblies, filled and unfilled market requirements, uranium imports and exports, and uranium inventories are shown in Chapter 2. A feature article, The Role of Thorium in Nuclear Energy, is included. 24 figs., 56 tabs.

The Uranium Industry Annual provides current statistical data on the US uranium industry for the Congress, Federal and State agencies, the uranium and electric utility industries, and the public. The feature article, ``Decommissioning of US Conventional Uranium Production Centers,`` is included. Data on uranium raw materials activities including exploration activities and expenditures, resources and reserves, mine production of uranium, production of uranium concentrate, and industry employment are presented in Chapter 1. Data on uranium marketing activities including domestic uranium purchases, commitments by utilities, procurement arrangements, uranium imports under purchase contracts and exports, deliveries to enrichment suppliers, inventories, secondary market activities, utility market requirements, and uranium for sale by domestic suppliers are presented in Chapter 2.

Depleted uranium (DU) is produced as a by-product of the uranium enrichment process. Over 340,000 MTU of DU in the form of UF{sub 6} have been accumulated at the US government gaseous diffusion plants and the stockpile continues to grow. An overview of issues and objectives associated with the inventory management and the ultimate disposition of this material is presented.

about 2.75 wt% U-235. For further enrichment, the material was shipped to the Oak Ridge and Portsmouth plants. In addition to natural uranium, also uranium recycled from spent fuel was fed into the Paducah enrichment cascade (Table 2 and Fig. 2). The recycled uranium introduced various isotopes not found in natural uranium into the cascade: fission products, such as Technetium-99; transuranics, such as Neptunium-237 and Plutonium-239; and the artificial uranium isotope of Uranium-236. The spent fuel, from which uranium was recycled, originated from the Hanford and Savannah River military plutonium production reactors. This uranium was recycled, although its assay of U-235 was somewhat lower than in natural uranium (Table 2). This obviously must be seen in the context of the Cold War era, when uranium was a scarce resource. Due to the low burn-up of the military reactors, concentrations of artificial U-236 are comparatively low in this recycled uranium. The recycled uranium represents

A new process for the extraction of uranium from seawater was developed. In the process, uranium adsorption is effected using powdered magnetic adsorbents; the adsorbents are then separated from seawater using magnetic separation technology. This process is superior to a column method using a granulated hydrous titanium oxide adsorber bed in the following ways: (1) a higher rate of adsorption is realized because smaller particles are used in the uranium adsorption; and (2) blocking, which is inevitable in an adsorber bed, is eliminated. The composite hydrous titanium-iron oxide as a magnetic adsorbent having high uranium adsorption capacity and magnetization can be prepared by adding urea to a mixed solution of titanium sulfate and ferrous sulfate. Adsorption and desoprtion of uranium and the removal of the adsorbent using a small-scale uranium extraction plant (about 15 m/sup 3//d) is reported, and the feasibility of uranium extraction from seawater by this process is demonstrated. 10 figures.

and two beta-minus decays of intermediate nuclei. Therefore plutonium production requires irradiation of natural or depleted uranium with a high neutron flux and chemical separation of plutonium from uranium and fission products. 3. Similar... Nuclear Facilities and Site Signatures.................................. 45 III.C.1. Uranium from Mining Operations........................................... 48 III.C.2. Uranium by inSituLeaching...

The Uranium Industry Annual 1998 (UIA 1998) provides current statistical data on the US uranium industry`s activities relating to uranium raw materials and uranium marketing. It contains data for the period 1989 through 2008 as collected on the Form EIA-858, ``Uranium Industry Annual Survey.`` Data provides a comprehensive statistical characterization of the industry`s activities for the survey year and also include some information about industry`s plans and commitments for the near-term future. Data on uranium raw materials activities for 1989 through 1998, including exploration activities and expenditures, EIA-estimated reserves, mine production of uranium, production of uranium concentrate, and industry employment, are presented in Chapter 1. Data on uranium marketing activities for 1994 through 2008, including purchases of uranium and enrichment services, enrichment feed deliveries, uranium fuel assemblies, filled and unfilled market requirements, and uranium inventories, are shown in Chapter 2. The methodology used in the 1998 survey, including data edit and analysis, is described in Appendix A. The methodologies for estimation of resources and reserves are described in Appendix B. A list of respondents to the ``Uranium Industry Annual Survey`` is provided in Appendix C. The Form EIA-858 ``Uranium Industry Annual Survey`` is shown in Appendix D. For the readers convenience, metric versions of selected tables from Chapters 1 and 2 are presented in Appendix E along with the standard conversion factors used. A glossary of technical terms is at the end of the report. 24 figs., 56 tabs.

The Uranium Industry Annual 1994 (UIA 1994) provides current statistical data on the US uranium industry`s activities relating to uranium raw materials and uranium marketing during that survey year. The UIA 1994 is prepared for use by the Congress, Federal and State agencies, the uranium and nuclear electric utility industries, and the public. It contains data for the 10-year period 1985 through 1994 as collected on the Form EIA-858, ``Uranium Industry Annual Survey.`` Data collected on the ``Uranium Industry Annual Survey`` (UIAS) provide a comprehensive statistical characterization of the industry`s activities for the survey year and also include some information about industry`s plans and commitments for the near-term future. Where aggregate data are presented in the UIA 1994, care has been taken to protect the confidentiality of company-specific information while still conveying accurate and complete statistical data. A feature article, ``Comparison of Uranium Mill Tailings Reclamation in the United States and Canada,`` is included in the UIA 1994. Data on uranium raw materials activities including exploration activities and expenditures, EIA-estimated resources and reserves, mine production of uranium, production of uranium concentrate, and industry employment are presented in Chapter 1. Data on uranium marketing activities, including purchases of uranium and enrichment services, and uranium inventories, enrichment feed deliveries (actual and projected), and unfilled market requirements are shown in Chapter 2.

To address the management of uranium-contaminated soils at Fernald and other DOE sites, the DOE Office of Technology Development formed the Uranium in Soils Integrated Demonstration (USID) program. The USID has five major tasks. These include the development and demonstration of technologies that are able to (1) characterize the uranium in soil, (2) decontaminate or remove uranium from the soil, (3) treat the soil and dispose of any waste, (4) establish performance assessments, and (5) meet necessary state and federal regulations. This report deals with soil decontamination or removal of uranium from contaminated soils. The report was compiled by the USID task group that addresses soil decontamination; includes data from projects under the management of four DOE facilities [Argonne National Laboratory (ANL), Los Alamos National Laboratory (LANL), Oak Ridge National Laboratory (ORNL), and the Savannah River Plant (SRP)]; and consists of four separate reports written by staff at these facilities. The fundamental goal of the soil decontamination task group has been the selective extraction/leaching or removal of uranium from soil faster, cheaper, and safer than current conventional technologies. The objective is to selectively remove uranium from soil without seriously degrading the soil`s physicochemical characteristics or generating waste forms that are difficult to manage and/or dispose of. Emphasis in research was placed more strongly on chemical extraction techniques than physical extraction techniques.

The operating uranium enrichment gaseous diffusion plants (GDPs) in Portsmouth, Ohio and Paducah, Kentucky, which are operated for the United States Department for Energy by Martin Marietta Energy Systems (MMES), currently use a chlorofluorocarbon (CFC-114) as the primary process stream coolant. Due to recent legislation embodied in the Clean Air Act, the production of this and other related chlorofluorocarbons (CFCS) are to be phased out with no production occurring after 1995. Since the plants lose approximately 500,000 pounds per year of this process stream coolant through various leaks, the GDPs are faced with the challenge of identifying a replacement coolant that will allow continued operation of the plants. MMES formed the CFC Task Team to identify and solve the various problems associated with identifying and implementing a replacement coolant. This report includes a review of the work performed by the CFC Task Team, and recommendations that were formulated based on this review and upon original work. The topics covered include; identifying a replacement coolant, coolant leak detection and repair efforts, coolant safety concerns, coolant level sensors, regulatory issues, and an analytical decision analysis.

The Uranium Industry Annual 1995 (UIA 1995) provides current statistical data on the U.S. uranium industry`s activities relating to uranium raw materials and uranium marketing. The UIA 1995 is prepared for use by the Congress, Federal and State agencies, the uranium and nuclear electric utility industries, and the public. It contains data for the period 1986 through 2005 as collected on the Form EIA-858, ``Uranium Industry Annual Survey``. Data collected on the ``Uranium Industry Annual Survey`` provide a comprehensive statistical characterization of the industry`s plans and commitments for the near-term future. Where aggregate data are presented in the UIA 1995, care has been taken to protect the confidentiality of company-specific information while still conveying accurate and complete statistical data. Data on uranium raw materials activities for 1986 through 1995 including exploration activities and expenditures, EIA-estimated reserves, mine production of uranium, production of uranium concentrate, and industry employment are presented in Chapter 1. Data on uranium marketing activities for 1994 through 2005, including purchases of uranium and enrichment services, enrichment feed deliveries, uranium fuel assemblies, filled and unfilled market requirements, uranium imports and exports, and uranium inventories are shown in Chapter 2. The methodology used in the 1995 survey, including data edit and analysis, is described in Appendix A. The methodologies for estimation of resources and reserves are described in Appendix B. A list of respondents to the ``Uranium Industry Annual Survey`` is provided in Appendix C. For the reader`s convenience, metric versions of selected tables from Chapters 1 and 2 are presented in Appendix D along with the standard conversion factors used. A glossary of technical terms is at the end of the report. 14 figs., 56 tabs.

Soils and sediments contaminated with uranium pose certain environmental and ecological risks. At low to moderate levels of contamination, the magnitude of these risks depends not only on the absolute concentrations of uranium in the material but also on the availability of the uranium to drinking water supplies, plants, or higher organisms. Rational approaches for regulating the clean-up of sites contaminated with uranium, therefore, should consider the value of assessing the environmental availability of uranium at the site before making decisions regarding remediation. The purpose of this work is to review existing approaches and procedures to determine their potential applicability for assessing the environmental availability of uranium in bulk soils or sediments. In addition to making the recommendations regarding methodology, the authors have tabulated data from the literature on the aqueous complexes of uranium and major uranium minerals, examined the possibility of predicting environmental availability of uranium based on thermodynamic solubility data, and compiled a representative list of analytical laboratories capable of performing environmental analyses of uranium in soils and sediments.

be discharged to water. Radioactive Wastes--Wastes managed for their radioactive content. Spent Nuclear Fuels--Fuel plants with reactors that use water for moderating nuclear reactions and cooling. Spent Nuclear Fuel Used or"spent"nuclear fuel is stored in pools, or in specially designed dry storage casks. Fabrication

A process is described for converting scrap and waste uranium oxide to uranium metal. The uranium oxide is sequentially reduced with a suitable reducing agent to a mixture of uranium metal and oxide products. The uranium metal is then converted to uranium hydride and the uranium hydride-containing mixture is then cooled to a temperature less than -100.degree. C. in an inert liquid which renders the uranium hydride ferromagnetic. The uranium hydride is then magnetically separated from the cooled mixture. The separated uranium hydride is readily converted to uranium metal by heating in an inert atmosphere. This process is environmentally acceptable and eliminates the use of hydrogen fluoride as well as the explosive conditions encountered in the previously employed bomb-reduction processes utilized for converting uranium oxides to uranium metal.

The U.S. Department of Energy (DOE) has a management challenge and financial liability in the form of 50,000 cylinders containing 555,000 metric tons of depleted uranium hexafluoride (UF{sub 6}) that are stored at the gaseous diffusion plants. The annual storage and maintenance cost is approximately $10 million. This report summarizes several studies undertaken by the DOE Office of Technology Development (OTD) to evaluate options for long-term depleted uranium management. Based on studies conducted to date, the most likely use of the depleted uranium is for shielding of spent nuclear fuel (SNF) or vitrified high-level waste (HLW) containers. The alternative to finding a use for the depleted uranium is disposal as a radioactive waste. Estimated disposal costs, utilizing existing technologies, range between $3.8 and $11.3 billion, depending on factors such as applicability of the Resource Conservation and Recovery Act (RCRA) and the location of the disposal site. The cost of recycling the depleted uranium in a concrete based shielding in SNF/HLW containers, although substantial, is comparable to or less than the cost of disposal. Consequently, the case can be made that if DOE invests in developing depleted uranium shielded containers instead of disposal, a long-term solution to the UF{sub 6} problem is attained at comparable or lower cost than disposal as a waste. Two concepts for depleted uranium storage casks were considered in these studies. The first is based on standard fabrication concepts previously developed for depleted uranium metal. The second converts the UF{sub 6} to an oxide aggregate that is used in concrete to make dry storage casks.

The limiting value for uranium toxicity in a human being should be based on the concentration of uranium (U) in the kidneys. The threshold for nephrotoxicity appears to lie very near 3 {mu}g U per gram kidney tissue. There does not appear to be strong scientific support for any other improved estimate, either higher or lower than this, of the threshold for uranium nephrotoxicity in a human being. The value 3 {mu}g U per gram kidney is the concentration that results from a single intake of about 30 mg soluble uranium by inhalation (assuming the metabolism of a standard person). The concentration of uranium continues to increase in the kidneys after long-term, continuous (or chronic) exposure. After chronic intakes of soluble uranium by workers at the rate of 10 mg U per week, the concentration of uranium in the kidneys approaches and may even exceed the nephrotoxic limit of 3 {mu}g U per gram kidney tissue. Precise values of the kidney concentration depend on the biokinetic model and model parameters assumed for such a calculation. Since it is possible for the concentration of uranium in the kidneys to exceed 3 {mu}g per gram tissue at an intake rate of 10 mg U per week over long periods of time, we believe that the kidneys are protected from injury when intakes of soluble uranium at the rate of 10 mg U per week do not continue for more than two consecutive weeks. For long-term, continuous occupational exposure to low-level, soluble uranium, we recommend a reduced weekly intake limit of 5 mg uranium to prevent nephrotoxicity in workers. Our analysis shows that the nephrotoxic limit of 3 {mu}g U per gram kidney tissues is not exceeded after long-term, continuous uranium intake at the intake rate of 5 mg soluble uranium per week.

The results of reactor tests of a carbonitride fuel in a single-crystal cladding from a molybdenum-based alloy can be used in substantiating the operational reliability of fuels in developing a project of a megawatt space nuclear power plant. The results of experimental studies of uranium and cesium penetration into the single-crystal cladding of fuel elements with a carbonitride fuel are interpreted. Those fuel elements passed nuclear power tests in the Ya-82 pilot plant for 8300 h at a temperature of about 1500°C. It is shown that the diffusion coefficients for uranium diffusion into the cladding are virtually coincident with the diffusion coefficients measured earlier for uranium diffusion into polycrystalline molybdenum. It is found that the penetration of uranium into the cladding is likely to occur only in the case of a direct contact between the cladding and fuel. The experimentally observed nonmonotonic uranium-concentration profiles are explained in terms of predominant uranium diffusion along grain boundaries. It is shown that a substantially nonmonotonic behavior observed in our experiment for the uranium-concentration profile may be explained by the presence of a polycrystalline structure of the cladding in the surface region from its inner side. The diffusion coefficient is estimated for the grain-boundary diffusion of uranium. The diffusion coefficients for cesium are estimated on the basis of experimental data obtained in the present study.

A method is described for forming metallic uranium, or a uranium alloy, from uranium oxide in a manner which substantially eliminates the formation of uranium-containing wastes. A source of uranium dioxide is first provided, for example, by reducing uranium trioxide (UO{sub 3}), or any other substantially stable uranium oxide, to form the uranium dioxide (UO{sub 2}). This uranium dioxide is then chlorinated to form uranium tetrachloride (UCl{sub 4}), and the uranium tetrachloride is then reduced to metallic uranium by reacting the uranium chloride with a metal which will form the chloride of the metal. This last step may be carried out in the presence of another metal capable of forming one or more alloys with metallic uranium to thereby lower the melting point of the reduced uranium product. The metal chloride formed during the uranium tetrachloride reduction step may then be reduced in an electrolysis cell to recover and recycle the metal back to the uranium tetrachloride reduction operation and the chlorine gas back to the uranium dioxide chlorination operation. 4 figs.

A method is described for forming metallic uranium, or a uranium alloy, from uranium oxide in a manner which substantially eliminates the formation of uranium-containing wastes. A source of uranium dioxide is first provided, for example, by reducing uranium trioxide (UO.sub.3), or any other substantially stable uranium oxide, to form the uranium dioxide (UO.sub.2). This uranium dioxide is then chlorinated to form uranium tetrachloride (UCl.sub.4), and the uranium tetrachloride is then reduced to metallic uranium by reacting the uranium chloride with a metal which will form the chloride of the metal. This last step may be carried out in the presence of another metal capable of forming one or more alloys with metallic uranium to thereby lower the melting point of the reduced uranium product. The metal chloride formed during the uranium tetrachloride reduction step may then be reduced in an electrolysis cell to recover and recycle the metal back to the uranium tetrachloride reduction operation and the chlorine gas back to the uranium dioxide chlorination operation.

UI.sub.3(1,4-dioxane).sub.1.5 and UI.sub.4(1,4-dioxane).sub.2, were synthesized in high yield by reacting turnings of elemental uranium with iodine dissolved in 1,4-dioxane under mild conditions. These molecular compounds of uranium are thermally stable and excellent precursor materials for synthesizing other molecular compounds of uranium including alkoxide, amide, organometallic, and halide compounds.

After almost 40 years of operation, the federal government is withdrawing from the uranium enrichment business. On July 1, the Department of Energy turned over to a new government-owned entity--the US Enrichment Corp. (USEC)--both the DOE enrichment plants at Paducah, Ky., and Portsmouth, Ohio, and domestic and international marketing of enriched uranium from them. Pushed by the inability of DOE's enrichment operations to meet foreign competition, Congress established USEC under the National Energy Policy Act of 1992, envisioning the new corporation as the first step to full privatization. With gross revenues of $1.5 billion in fiscal 1992, USEC would rank 275th on the Fortune 500 list of top US companies. USEC will lease from DOE the Paducah and Portsmouth facilities, built in the early 1950s, which use the gaseous diffusion process for uranium enrichment. USEC's stock is held by the US Treasury, to which it will pay annual dividends. Martin Marietta Energy Systems, which has operated Paducah since 1984 and Portsmouth since 1986 for DOE, will continue to operate both plants for USEC. Closing one of the two facilities will be studied, especially in light of a 40% world surplus of capacity over demand. USEC also will consider other nuclear-fuel-related ventures. USEC will produce only low-enriched uranium, not weapons-grade material. Indeed, USEC will implement a contract now being completed under which the US will purchase weapons-grade uranium from dismantled Russian nuclear weapons and convert it into low-enriched uranium for power reactor fuel.

The United States Government currently has an abundance of depleted uranium (DU). This surplus of about 1 billion pounds is the result of an enrichment process using gaseous diffusion to produce enriched and depleted uranium. The enriched uranium has been used primarily for either nuclear weapons for the military or nuclear fuel for the commercial power industry. Most of the depleted uranium remains at the enrichment process plants in the form of depleted uranium hexafluoride (DUF{sub 6}). The Department of Energy (DOE) recently began a study to identify possible commercial applications for the surplus material. One of these potential applications is to use the DU in high-density strikers/hammers in pneumatically driven tools, such as jack hammers and piledrivers to improve their impulse performance. The use of DU could potentially increase tunneling velocity and excavation into target materials with improved efficiency. This report describes the efforts undertaken to analyze the particulars of using DU in two specific striking applications: the jackhammer and chipper tool.

This is a single stage process for treating spent nuclear fuel from light water reactors. The spent nuclear fuel, uranium oxide, UO.sub.2, is added to a solution of UCl.sub.4 dissolved in molten LiCl. A carbon anode and a metallic cathode is positioned in the molten salt bath. A power source is connected to the electrodes and a voltage greater than or equal to 1.3 volts is applied to the bath. At the anode, the carbon is oxidized to form carbon dioxide and uranium chloride. At the cathode, uranium is electroplated. The uranium chloride at the cathode reacts with more uranium oxide to continue the reaction. The process may also be used with other transuranic oxides and rare earth metal oxides.

t in the depleted uranium. For this purpose, we first need to calculate the mass balance of the enrichment process. We then calculate the inhalation doses from the depleted uranium and compare the dose contributions from the nuclides of interest. Mass balance for uranium enrichment at Paducah [DOE_1984, p.35] Feed Product Tails Other Mass [st] 758002 124718 621894 11390 Mass fraction 100.00% 16.45% 82.04% 1.50% Concentration of plutonium in tails (depleted uranium) from enrichment of reprocessed uranium, assuming that all plutonium were transfered to the tails: Concentration of neptunium in tails from enrichment of reprocessed uraniumuranium, assuming that all neptunium were transfered to the tails: - 2 - Schematic of historic uranium enrichment process at Paducah [DOE_1999b] - -7 For comparison, we first calculate the inhalation dose from depleted uranium produced from natural uranium. We assume that the short-lived decay products have reached secular equilibrium with th

This report evaluates two management alternatives for Department of Energy depleted uranium: continued storage as uranium hexafluoride, and conversion to uranium metal and fabrication to shielding for spent nuclear fuel containers. The results will be used to compare the costs with other alternatives, such as disposal. Cost estimates for the continued storage alternative are based on a life-cycle of 27 years through the year 2020. Cost estimates for the recycle alternative are based on existing conversion process costs and Capital costs for fabricating the containers. Additionally, the recycle alternative accounts for costs associated with intermediate product resale and secondary waste disposal for materials generated during the conversion process.

Current liquid-liquid extraction processes used in recycling irradiated nuclear fuel rely on (1) strong nitric acid to dissolve uranium oxide fuel, and (2) the use of aliphatic hydrocarbons as a diluent in formulating the solvent used to extract uranium. The nitric acid dissolution process is not selective. It dissolves virtually the entire fuel meat which complicates the uranium extraction process. In addition, a solvent washing process is used to remove TBP degradation products, which adds complexity to the recycling plant and increases the overall plant footprint and cost. A liquid or supercritical carbon dioxide (l/sc -CO2) system was designed to mitigate these problems. Indeed, TBP nitric acid complexes are highly soluble in l/sc -CO2 and are capable of extracting uranium directly from UO2, UO3 and U3O8 powders. This eliminates the need for total acid dissolution of the irradiated fuel. Furthermore, since CO2 is easily recycled by evaporation at room temperature and pressure, it eliminates the complex solvent washing process. In this report, we demonstrate: (1) A reprocessing scheme starting with the selective extraction of uranium from solid uranium oxides into a TBP-HNO3 loaded Sc-CO2 phase, (2) Back extraction of uranium into an aqueous phase, and (3) Conversion of recovered purified uranium into uranium oxide. The purified uranium product from step 3 can be disposed of as low level waste, or mixed with enriched uranium for use in a reactor for another fuel cycle. After an introduction on the concept and properties of supercritical fluids, we first report the characterization of the different oxides used for this project. Our extraction system and our online monitoring capability using UV-Vis absorbance spectroscopy directly in sc-CO2 is then presented. Next, the uranium extraction efficiencies and kinetics is demonstrated for different oxides and under different physical and chemical conditions: l/sc -CO2 pressure and temperature, TBP/HNO3 complex used, reductant or complexant used for selectivity, and ionic liquids used as supportive media. To complete the extraction and recovery cycle, we then demonstrate uranium back extraction from the TBP loaded sc-CO2 phase into an aqueous phase and the characterization of the uranium complex formed at the end of this process. Another aspect of this project was to limit proliferation risks by either co-extracting uranium and plutonium, or by leaving plutonium behind by selectively extracting uranium. We report that the former is easily achieved, since plutonium is in the tetravalent or hexavalent oxidation state in the oxidizing environment created by the TBP-nitric acid complex, and is therefore co-extracted. The latter is more challenging, as a reductant or complexant to plutonium has to be used to selectively extract uranium. After undertaking experiments on different reducing or complexing systems (e.g., AcetoHydroxamic Acid (AHA), Fe(II), ascorbic acid), oxalic acid was chosen as it can complex tetravalent actinides (Pu, Np, Th) in the aqueous phase while allowing the extraction of hexavalent uranium in the sc-CO2 phase. Finally, we show results using an alternative media to commonly used aqueous phases: ionic liquids. We show the dissolution of uranium in ionic liquids and its extraction using sc-CO2 with and without the presence of AHA. The possible separation of trivalent actinides from uranium is also demonstrated in ionic liquids using neodymium as a surrogate and diglycolamides as the extractant.

The invention is a novel method for the recovery of uranium from dry, particulate uranium tetrafluoride. In one aspect, the invention comprises reacting particulate uranium tetrafluoride and calcium oxide in the presence of gaseous oxygen to effect formation of the corresponding alkaline earth metal uranate and alkaline earth metal fluoride. The product uranate is highly soluble in various acidic solutions wherein the product fluoride is virtually insoluble therein. The product mixture of uranate and alkaline earth metal fluoride is contacted with a suitable acid to provide a uranium-containing solution, from which the uranium is recovered. The invention can achieve quantitative recovery of uranium in highly pure form.

The invention is a novel method for the recovery of uranium from dry, particulate uranium tetrafluoride. In one aspect, the invention comprises reacting particulate uranium tetrafluoride and calcium oxide in the presence of gaseous oxygen to effect formation of the corresponding alkaline earth metal uranate and alkaline earth metal fluoride. The product uranate is highly soluble in various acidic solutions whereas the product fluoride is virtually insoluble therein. The product mixture of uranate and alkaline earth metal fluoride is contacted with a suitable acid to provide a uranium-containing solution, from which the uranium is recovered. The invention can achieve quantitative recovery of uranium in highly pure form.

From 1980 to 1993, the domestic production of uranium declined from almost 44 million pounds U{sub 3}O{sub 8} to about 3 million pounds. This retrenchment of the U.S. uranium industry resulted in the permanent closing of many uranium-producing facilities. Current low uranium prices, excess world supply, and low expectations for future uranium demand indicate that it is unlikely existing plants will be reopened. Because of this situation, these facilities eventually will have to be decommissioned. The Uranium Mill Tailings and Radiation Control Act of 1978 (UMTRCA) vests the U.S. Environmental Protection Agency (EPA) with overall responsibility for establishing environmental standards for decommissioning of uranium production facilities. UMTRCA also gave the U.S. Nuclear Regulatory Commission (NRC) the responsibility for licensing and regulating uranium production and related activities, including decommissioning. Because there are many issues associated with decommissioning-environmental, political, and financial-this report will concentrate on the answers to three questions: (1) What is required? (2) How is the process implemented? (3) What are the costs? Regulatory control is exercised principally through the NRC licensing process. Before receiving a license to construct and operate an uranium producing facility, the applicant is required to present a decommissioning plan to the NRC. Once the plan is approved, the licensee must post a surety to guarantee that funds will be available to execute the plan and reclaim the site. This report by the Energy Information Administration (EIA) represents the most comprehensive study on this topic by analyzing data on 33 (out of 43) uranium production facilities located in Colorado, Nebraska, New Mexico, South Dakota, Texas, Utah, and Washington.

Lynas Advanced Materials Plant (LAMP) has been licensed to produce the rare earths elements since early 2013 in Malaysia. LAMP processes lanthanide concentrate (LC) to extract rare earth elements and subsequently produce large volumes of water leach purification (WLP) residue containing naturally occurring radioactive material (NORM). This residue has been rising up the environmental issue because it was suspected to accumulate thorium with significant activity concentration and has been classified as radioactive residue. The aim of this study is to determine Th-232, U-238 and rare earth elements in lanthanide concentrate (LC) and water leach purification (WLP) residue collected from LAMP and to evaluate the potential radiological impacts of the WLP residue on the environment. Instrumental Neutron Activation Analysis and ?-spectrometry were used for determination of Th, U and rare earth elements concentrations. The results of this study found that the concentration of Th in LC was 1289.7 ± 129 ppm (5274.9 ± 527.6Bq/kg) whereas the Th and U concentrations in WLP were determined to be 1952.9±17.6 ppm (7987.4 ± 71.9 Bq/kg) and 17.2 ± 2.4 ppm respectively. The concentrations of Th and U in LC and WLP samples determined by ?- spectrometry were 1156 ppm (4728 ± 22 Bq/kg) and 18.8 ppm and 1763.2 ppm (7211.4 Bq/kg) and 29.97 ppm respectively. This study showed that thorium concentrations were higher in WLP compare to LC. This study also indicate that WLP residue has high radioactivity of {sup 232}Th compared to Malaysian soil natural background (63 - 110 Bq/kg) and come under preview of Act 304 and regulations. In LC, the Ce and Nd concentrations determined by INAA were 13.2 ± 0.6% and 4.7 ± 0.1% respectively whereas the concentrations of La, Ce, Nd and Sm in WLP were 0.36 ± 0.04%, 1.6%, 0.22% and 0.06% respectively. This result showed that some amount of rare earth had not been extracted and remained in the WLP and may be considered to be reextracted.

plants (U.S. DOE/EIA 2003a, 2003b, 2006). Uranium exploration, mining, and ore processing left a legacy.S. uranium production spanned from approximately 1948 to the early 1980s (U.S. DOE/EIA 1992). Through 2005 the industry had generated over 420,000 metric tons (MTs) of uranium for nuclear weapons and commercial power

A process for making uranium metal from uranium oxide by first fluorinating uranium oxide to form uranium tetrafluoride and next electrolytically reducing the uranium tetrafluoride with a carbon anode to form uranium metal and CF.sub.4. The CF.sub.4 is reused in the fluorination reaction rather than being disposed of as a hazardous waste.

for the last decade. Most of their work involves depleted uranium, a more common form of uraniumMarch 2008 Controlling uranium reactivity March 18, 2008 Uranium is an often misunderstood metal uranium research. In reality, uranium presents a wealth of possibilities for funda- mental chemistry. Many

This work addresses concerns that the rapid, exothermic oxidation of active uranium hydride in air could stimulate an exothermic reaction (burning) involving any adjacent uranium metal, so as to increase the potential hazard arising from a hydride reaction. The effect of the thermal reaction of active uranium hydride, especially in contact with uranium metal, does not increase in proportion with hydride mass, particularly when considering large quantities of hydride. Whether uranium metal continues to burn in the long term is a function of the uranium metal and its surroundings. The source of the initial heat input to the uranium, if sufficient to cause ignition, is not important. Sustained burning of uranium requires the rate of heat generation to be sufficient to offset the total rate of heat loss so as to maintain an elevated temperature. For dense uranium, this is very difficult to achieve in naturally occurring circumstances. Areas of the uranium surface can lose heat but not generate heat. Heat can be lost by conduction, through contact with other materials, and by convection and radiation, e.g. from areas where the uranium surface is covered with a layer of oxidised material, such as burned-out hydride or from fuel cladding. These rates of heat loss are highly significant in relation to the rate of heat generation by sustained oxidation of uranium in air. Finite volume modelling has been used to examine the behaviour of a magnesium-clad uranium metal fuel element within a bottle surrounded by other un-bottled fuel elements. In the event that the bottle is breached, suddenly, in air, it can be concluded that the bulk uranium metal oxidation reaction will not reach a self-sustaining level and the mass of uranium oxidised will likely to be small in relation to mass of uranium hydride oxidised. (authors)

Although there are several secondary issues, the most important uranium resource issue is, ``will there be enough uranium available at a cost which will allow nuclear power to be competitive in the future?`` This paper will attempt to answer this question by discussing uranium supply, demand, and economics from the perspective of the United States. The paper will discuss: how much uranium is available; the sensitivity of nuclear power costs to uranium price; the potential future demand for uranium in the Unites States, some of the options available to reduce this demand, the potential role of the Advanced Liquid Metal Cooled Reactor (ALMR) in reducing uranium demand; and potential alternative uranium sources and technologies.

Brazil is a country of vast natural resources, including numerous uranium deposits. In support of the country`s nuclear power program, Brazil has developed the most active uranium industry in South America. Brazil has one operating reactor (Angra 1, a 626-MWe PWR), and two under construction. The country`s economic challenges have slowed the progress of its nuclear program. At present, the Pocos de Caldas district is the only active uranium production. In 1990, the Cercado open-pit mine produced approximately 45 metric tons (MT) U{sub 3}O{sub 8} (100 thousand pounds). Brazil`s state-owned uranium production and processing company, Uranio do Brasil, announced it has decided to begin shifting its production from the high-cost and nearly depleted deposits at Pocos de Caldas, to lower-cost reserves at Lagoa Real. Production at Lagoa Real is schedules to begin by 1993. In addition to these two districts, Brazil has many other known uranium deposits, and as a whole, it is estimated that Brazil has over 275,000 MT U{sub 3}O{sub 8} (600 million pounds U{sub 3}O{sub 8}) in reserves.

Considerable information useful in nuclear waste storage can be gained by studying the conditions of uranium ore deposit formation. Further information can be gained by comparing the chemistry of uranium to nuclear fission products and other radionuclides of concern to nuclear waste disposal. Redox state appears to be the most important variable in controlling uranium solubility, especially at near neutral pH, which is characteristic of most ground water. This is probably also true of neptunium, plutonium, and technetium. Further, redox conditions that immobilize uranium should immobilize these elements. The mechanisms that have produced uranium ore bodies in the Earth's crust are somewhat less clear. At the temperatures of hydrothermal uranium deposits, equilibrium models are probably adequate, aqueous uranium (VI) being reduced and precipitated by interaction with ferrous-iron-bearing oxides and silicates. In lower temperature roll-type uranium deposits, overall equilibrium may not have been achieved. The involvement of sulfate-reducing bacteria in ore-body formation has been postulated, but is uncertain. Reduced sulfur species do, however, appear to be involved in much of the low temperature uranium precipitation. Assessment of the possibility of uranium transport in natural ground water is complicated because the system is generally not in overall equilibrium. For this reason, Eh measurements are of limited value. If a ground water is to be capable of reducing uranium, it must contain ions capable of reducing uranium both thermodynamically and kinetically. At present, the best candidates are reduced sulfur species.

An assessment was performed to show the impact on airborne release fraction, respirable fraction, dose conversion factor and dose consequences of postulated accidents at the Plutonium Finishing Plant involving uranium dioxide rather than plutonium dioxide.

The present invention is directed to the protecting of uranium and uranium alloy articles from corrosion by providing the surfaces of the articles with a layer of an ion-plated metal selected from aluminum and zinc to a thickness of at least 60 microinches and then converting at least the outer surface of the ion-plated layer of aluminum or zinc to aluminum chromate or zinc chromate. This conversion of the aluminum or zinc to the chromate form considerably enhances the corrosion resistance of the ion plating so as to effectively protect the coated article from corrosion.

Two embodiments of a high uranium fuel plate are disclosed which contain a meat comprising structured uranium compound confined between a pair of diffusion bonded ductile metal cladding plates uniformly covering the meat, the meat having a uniform high fuel loading comprising a content of uranium compound greater than about 45 Vol. % at a porosity not greater than about 10 Vol. %. In a first embodiment, the meat is a plurality of parallel wires of uranium compound. In a second embodiment, the meat is a dispersion compact containing uranium compound. The fuel plates are fabricated by a hot isostatic pressing process.

The Paducah and Portsmouth Gaseous Diffusion Plants enrich uranium in the {sup 235}U isotope by diffusing gaseous uranium hexafluoride (UF{sub 6}) through a porous barrier. The UF{sub 6} gaseous diffusion cascade utilized several thousand {open_quotes}stages{close_quotes} of barrier to produce highly enriched uranium (HEU). Historically, Portsmouth has enriched the Paducah Gaseous Diffusion Plant`s product (typically 1.8 wt% {sup 235}U) as well as natural enrichment feed stock up to 97 wt%. Due to the chemical reactivity of UF{sub 6}, particularly with water, the formation of solid uranium deposits occur at a gaseous diffusion plant. Much of the equipment operates below atmospheric pressure, and deposits are formed when atmospheric air enters the cascade. Deposits may also be formed from UF{sub 6} reactions with oil, UF{sub 6} reactions with the metallic surfaces of equipment, and desublimation of UF{sub 6}. The major deposits form as a result of moist air in leakage due to failure of compressor casing flanges, blow-off plates, seals, expansion joint convolutions, and instrument lines. This report describes criticality concerns and deposit disposition.

1.1 This specification is for finished sintered gadolinium oxide-uranium dioxide pellets for use in light-water reactors. It applies to gadolinium oxide-uranium dioxide pellets containing uranium of any 235U concentration and any concentration of gadolinium oxide. 1.2 This specification recognizes the presence of reprocessed uranium in the fuel cycle and consequently defines isotopic limits for gadolinium oxide-uranium dioxide pellets made from commercial grade UO2. Such commercial grade UO2 is defined so that, regarding fuel design and manufacture, the product is essentially equivalent to that made from unirradiated uranium. UO2 falling outside these limits cannot necessarily be regarded as equivalent and may thus need special provisions at the fuel fabrication plant or in the fuel design. 1.3 This specification does not include (1) provisions for preventing criticality accidents or (2) requirements for health and safety. Observance of this specification does not relieve the user of the obligation to be aw...

An important requirement for the international safeguards community is the ability to determine the enrichment level of uranium in gas centrifuge enrichment plants and nuclear fuel fabrication facilities. This is essential to ensure that countries with nuclear nonproliferation commitments, such as States Party to the Nuclear Nonproliferation Treaty, are adhering to their obligations. However, current technologies to verify the uranium enrichment level in gas centrifuge enrichment plants or nuclear fuel fabrication facilities are technically challenging and resource-intensive. NNSA’s Office of Nonproliferation and International Security (NIS) supports the development, testing, and evaluation of future systems that will strengthen and sustain U.S. safeguards and security capabilities—in this case, by automating the monitoring of uranium enrichment in the entire inventory of a fuel fabrication facility. One such system is HEVA—hybrid enrichment verification array. This prototype was developed to provide an automated, nondestructive assay verification technology for uranium hexafluoride (UF6) cylinders at enrichment plants.

An important requirement for the international safeguards community is the ability to determine the enrichment level of uranium in gas centrifuge enrichment plants and nuclear fuel fabrication facilities. This is essential to ensure that countries with nuclear nonproliferation commitments, such as States Party to the Nuclear Nonproliferation Treaty, are adhering to their obligations. However, current technologies to verify the uranium enrichment level in gas centrifuge enrichment plants or nuclear fuel fabrication facilities are technically challenging and resource-intensive. NNSA’s Office of Nonproliferation and International Security (NIS) supports the development, testing, and evaluation of future systems that will strengthen and sustain U.S. safeguards and security capabilities—in this case, by automating the monitoring of uranium enrichment in the entire inventory of a fuel fabrication facility. One such system is HEVA—hybrid enrichment verification array. This prototype was developed to provide an automated, nondestructive assay verification technology for uranium hexafluoride (UF6) cylinders at enrichment plants.

An expected windfall to the US Treasury from the sale of the Energy Dept.'s commercial fuel enrichment facilities may evaporate in the next few weeks when the Clinton administration submits its fiscal 1994 budget proposal to Congress, according to congressional and administration officials. Under the Energy Policy Act of 1992, DOE is required to lease two uranium enrichment facilities, Portsmouth, Ohio, and Paducah, KY., to the government-owned US Enrichment Corp. (USEC) by July 1. Estimates by OMB and Treasury indicate a potential yearly payoff of $300 million from the government-owned company's sale of fuel for commercial reactors. Those two facilities use a process of gaseous diffusion to enrich uranium to about 3 percent for use as fuel in commercial power plants. DOE has contracts through at least 1996 to provide about 12 million separative work units (SWUs) yearly to US utilities and others world-wide. But under an agreement signed between the US and Russia last August, at least 10 metric tons, or 1.5 million SWUs, of low-enriched uranium (LEU) blended down from Russia warheads is expected to be delivered to the US starting in 1994. It could be sold at $50 to $60 per SWU, far below what DOE currently charges for its SWUs - $135 per SWU for 70 percent of the contract price and $90 per SWU for the remaining 30 percent.

A method of producing thin foils of uranium or an alloy. The uranium or alloy is cast as a plate or sheet having a thickness less than about 5 mm and thereafter cold rolled in one or more passes at substantially ambient temperatures until the uranium or alloy thereof is in the shape of a foil having a thickness less than about 1.0 mm. The uranium alloy includes one or more of Zr, Nb, Mo, Cr, Fe, Si, Ni, Cu or Al.

Seawater contains various elements in solution. Deuterium, lithium, and uranium are the important ingredients for energy application at present and in the future. This paper deals with the recovery of uranium from seawater, with emphasis on the development of an adsorbent with high selectivity and rate of adsorption for uranium. Polyacrylamidoxime chelating resins were synthesized from various co-polymers of acrylonitrile and cross-linking agents. The resulting resins with the chelating amidoxime group showed selective adsorption for uranium in seawater. The amount of uranium adsorbed from seawater at room temperature reached 3.2 mg/g resin after 180 days. Polyacrylamidoxime fiber, which was prepared from polyacrylonitrile fiber and hydroxylamine, showed a high rate of adsorption for uranium. The polyacrylamidoxime fiber conditioned with 1 M HC1 and 1 M NaOH adsorbed 4 mg U/g fiber from seawater in ten days. 9 figures, 6 tables.

URANIUM IN PREHISTORIC INDIAN POTTERY A Thesis by ERNEST WILLIAM FILBERTH Submitted to the Graduate College of Texas A&M University in partial fulfillment of the requirement for the degree of MASTER OF SCIENCE December 1976 Major Subject...: Chemistry URANIUM IN PREHISTORIC INDIAN POTTERY A Thesis by ERNEST WILLIAM FILBERTH Approved as to style and content by: (Chairman of Committee) (Head of Department) (Member) (Membe (Member) (Member) December 1976 ABSTRACT Uranium in Prehistoric...

Significant amounts of the depleted uranium (DU) created by past uranium enrichment activities have been sold, disposed of commercially, or utilized by defense programs. In recent years, however, the demand for DU has become quite small compared to quantities available, and within the US Department of Energy (DOE) there is concern for any risks and/or cost liabilities that might be associated with the ever-growing inventory of this material. As a result, Martin Marietta Energy Systems, Inc. (Energy Systems), was asked to review options and to develop a comprehensive plan for inventory management and the ultimate disposition of DU accumulated at the gaseous diffusion plants (GDPs). An Energy Systems task team, under the chairmanship of T. R. Lemons, was formed in late 1989 to provide advice and guidance for this task. This report reviews options and recommends actions and objectives in the management of working inventories of partially depleted feed (PDF) materials and for the ultimate disposition of fully depleted uranium (FDU). Actions that should be considered are as follows. (1) Inspect UF{sub 6} cylinders on a semiannual basis. (2) Upgrade cylinder maintenance and storage yards. (3) Convert FDU to U{sub 3}O{sub 8} for long-term storage or disposal. This will include provisions for partial recovery of costs to offset those associated with DU inventory management and the ultimate disposal of FDU. Another recommendation is to drop the term tails'' in favor of depleted uranium'' or DU'' because the tails'' label implies that it is waste.'' 13 refs.

A process for converting UF.sub.6 to a solid uranium compound such as UO.sub.2 and CaF. The UF.sub.6 vapor form is contacted with an aqueous solution of NH.sub.4 OH at a pH greater than 7 to precipitate at least some solid uranium values as a solid leaving an aqueous solution containing NH.sub.4 OH and NH.sub.4 F and remaining uranium values. The solid uranium values are separated from the aqueous solution of NH.sub.4 OH and NH.sub.4 F and remaining uranium values which is then diluted with additional water precipitating more uranium values as a solid leaving trace quantities of uranium in a dilute aqueous solution. The dilute aqueous solution is contacted with an ion-exchange resin to remove substantially all the uranium values from the dilute aqueous solution. The dilute solution being contacted with Ca(OH).sub.2 to precipitate CaF.sub.2 leaving dilute NH.sub.4 OH.

A method for monitoring the stack gases of a purge cascade of gaseous diffusion plant for uranium activity. A sample stream is taken from the stack gases and contacted with a volume of moisture-laden air for converting trace levels of uranium hexafluoride, if any, in the stack gases into particulate uranyl fluoride. A continuous strip of filter paper from a supply roll is passed through this sampling stream to intercept and gather any uranyl fluoride in the sampling stream. This filter paper is then passed by an alpha scintillation counting device where any radioactivity on the filter paper is sensed so as to provide a continuous monitoring of the gas stream for activity indicative of the uranium content in the stack gases. 1 fig.

A method for monitoring the stack gases of a purge cascade of a gaseous diffusion plant for uranium activity. A sample stream is taken from the stack gases and contacted with a volume of moisture-laden air for converting trace levels of uranium hexafluoride, if any, in the stack gases into particulate uranyl fluoride. A continuous strip of filter paper from a supply roll is passed through this sampling stream to intercept and gather any uranyl fluoride in the sampling stream. This filter paper is then passed by an alpha scintillation counting device where any radioactivity on the filter paper is sensed so as to provide a continuous monitoring of the gas stream for activity indicative of the uranium content in the stack gases.

A Depleted Uranium Silicate COntainer Backfill System (DUSCOBS) is proposed that would use small, isotopically-depleted uranium silicate glass beads as a backfill material inside storage, transport, and repository waste packages containing spent nuclear fuel (SNF). The uranium silicate glass beads would fill all void space inside the package including the coolant channels inside SNF assemblies. Based on preliminary analysis, the following benefits have been identified. DUSCOBS improves repository waste package performance by three mechanisms. First, it reduces the radionuclide releases from SNF when water enters the waste package by creating a local uranium silicate saturated groundwater environment that suppresses (1) the dissolution and/or transformation of uranium dioxide fuel pellets and, hence, (2) the release of radionuclides incorporated into the SNF pellets. Second, the potential for long-term nuclear criticality is reduced by isotopic exchange of enriched uranium in SNF with the depleted uranium (DU) in the glass. Third, the backfill reduces radiation interactions between SNF and the local environment (package and local geology) and thus reduces generation of hydrogen, acids, and other chemicals that degrade the waste package system. In addition, the DUSCOBS improves the integrity of the package by acting as a packing material and ensures criticality control for the package during SNF storage and transport. Finally, DUSCOBS provides a potential method to dispose of significant quantities of excess DU from uranium enrichment plants at potential economic savings. DUSCOBS is a new concept. Consequently, the concept has not been optimized or demonstrated in laboratory experiments.

APPENDIX J Partition Coefficients For Uranium #12;Appendix J Partition Coefficients For Uranium J.1.0 Background The review of uranium Kd values obtained for a number of soils, crushed rock and their effects on uranium adsorption on soils are discussed below. The solution pH was also used as the basis

involve the use of natural or depleted uranium. Natural isotopes of uranium are U-238, U-235 and U-234 (see Table 1 for natural abundances). Depleted uranium contains less of the isotopes: U-235 and U-234. The specific activity of depleted uranium (5.0E-7 Ci/g) is less than that of natural uranium (7.1E-7 Ci

A recent effort determined uranium holdup at a large fuel fabrication facility abroad where low enriched ({approx} 3%) uranium (LEU) oxide feeds the pellet manufacturing process. Measurements taken with both high- and low-resolution gamma-ray spectrometry systems include extensive data for the ventilation and vacuum systems. Equipment dimensions and the corresponding holdup deposit masses are large for LEU. Because deposits are infinitely thick to the 186 keV gamma ray in many locations in an LEU environment, measurements of both the 186 and 1001 keV gamma-rays were required, and self-attenuation was significant at 1001 keV in many cases. These wide-dynamic-range measruements used short count times, portable scintillator detectors, and portable MCAs. Because equipment is elevated above floor levels, most measurements were made with detectors mounted on extended telescoping poles. One of the main goals of this effort was to demonstrate and validate methods for measurement and quantitative analysis of LEU holdup using low-resolution detectors and the Generalized Geometry Holdup (GGH) techniques. The current GGH approach is applied elsewhere for holdup measurements of plutonium and high-enriched uranium. The recent experience is directly applicable to holdup measruements at LEU facilities such as the Paducah and Portmouth gaseous diffusion enrichment plants and elsewhere, including LEU sites where D and D is active. This report discusses the measurement methodology, calibration of the measurement equipment, measurement control, analysis of the data, and the global and local assay results including random and systematic uncertainties. It includes field-validation exercises (multiple calibrated systems that perform measruements on the same extended equipment) as well as quantitative validation results obtained on reference materials assembled to emulate the deposits in an extended vacuum line that was also measured by these techniques. The paper examines the differences in assay results between the low-resolution system using the GGH method and the high-resolution system utilizing the commercially available ISOCS analysis method.

The U.S. Department of Energy (DOE) has a management challenge and financial liability problem in the form of 50,000 cylinders containing 555,000 metric tons of depleted uranium hexafluoride (UF{sub 6}) that are stored at the gaseous diffusion plants. DOE is evaluating several options for the disposition of this UF{sub 6}, including continued storage, disposal, and recycle into a product. Based on studies conducted to date, the most feasible recycle option for the depleted uranium is shielding in low-level waste, spent nuclear fuel, or vitrified high-level waste containers. Estimates for the cost of disposal, using existing technologies, range between $3.8 and $11.3 billion depending on factors such as the disposal site and the applicability of the Resource Conservation and Recovery Act (RCRA). Advanced technologies can reduce these costs, but UF{sub 6} disposal still represents large future costs. This paper describes an application for depleted uranium in which depleted uranium hexafluoride is converted into an oxide and then into a heavy aggregate. The heavy uranium aggregate is combined with conventional concrete materials to form an ultra high density concrete, DUCRETE, weighing more than 400 lb/ft{sup 3}. DUCRETE can be used as shielding in spent nuclear fuel/high-level waste casks at a cost comparable to the lower of the disposal cost estimates. Consequently, the case can be made that DUCRETE shielded casks are an alternative to disposal. In this case, a beneficial long term solution is attained for much less than the combined cost of independently providing shielded casks and disposing of the depleted uranium. Furthermore, if disposal is avoided, the political problems associated with selection of a disposal location are also avoided. Other studies have also shown cost benefits for low level waste shielded disposal containers.

A computer program entitled URPE (Uranium Recovery Performance and Economics) has been developed to simulate the engineering performance and provide an economic analysis of a plant recovering uranium from seawater. The conceptual system design used as the focal point for the more general analysis consists of a floating oil-rig type of platform single-point moored in an open ocean current, using either high-volume-low-head axial pumps or the velocity head of the ambient ocean current to force seawater through a mass transfer medium (hydrous titanium oxide (HTO) coated onto particle beds or stacked tubes). Uranium is recovered from the seawater by an adsorption process, and later eluted from the adsober by an ammonium carbonate solution. A multiproduct cogenerating plant on board the platform burns coal to raise steam for electricity generation, desalination, and process heat requirements. Scrubbed stack gas from the plant is processed to recover carbon dioxide for chemical make-up needs. The equilibrium isotherm and the diffusion constant for the uranyl-HTO system, which are needed for bed performance calculations, have been calculated based on the data reported in the literature. In addition, a technique for calculating the rate constant of a fixed-bed adsoorbing system has been developed for use with Thomas' solution for predicting fixed-bed performance.

A method is described for measuring the uranium content of aqueous solutions wherein a uranyl phosphate complex is irradiated with a 5 nanosecond pulse of 425 nanometer laser light and resultant 520 nanometer emissions are observed for a period of 50 to 400 microseconds after the pulse. Plotting the natural logarithm of emission intensity as a function of time yields an intercept value which is proportional to uranium concentration.

A method is described for measuring the uranium content of aqueous solutions wherein a uranyl phosphate complex is irradiated with a 5 nanosecond pulse of 425 nanometer laser light and resultant 520 nanometer emissions are observed for a period of 50 to 400 microseconds after the pulse. Plotting the natural logarithm of emission intensity as a function of time yields an intercept value which is proportional to uranium concentration.

Naturally occurring uranium contains 0.71 wt% {sup 235}U. In order for the uranium to be useful in most fission reactors, it must be enriched the concentration of the fissile isotope {sup 235}U must be increased. Depleted uranium (DU) is a co-product of the processing of natural uranium to produce enriched uranium, and DU has a {sup 235}U concentration of less than 0.71 wt%. In the United States, essentially all of the DU inventory is in the chemical form of uranium hexafluoride (UF{sub 6}) and is stored in large cylinders above ground. If this co-product material were to be declared surplus, converted to a stable oxide form, and disposed, the costs are estimated to be several billion dollars. Only small amounts of DU have at this time been beneficially reused. The U.S. Department of Energy (DOE) has begun the Beneficial Uses of DU Project to identify large-scale uses of DU and encourage its reuse for the primary purpose of potentially reducing the cost and expediting the disposition of the DU inventory. This paper discusses the inventory of DU and its rate of increase; DU disposition options; beneficial use options; a preliminary cost analysis; and major technical, institutional, and regulatory issues to be resolved.

Alloys such as U-6Nb are prepared by forming a stacked sandwich array of uranium sheets and niobium powder disposed in layers between the sheets, heating the array in a vacuum induction melting furnace to a temperature such as to melt the uranium, holding the resulting mixture at a temperature above the melting point of uranium until the niobium dissolves in the uranium, and casting the uranium-niobium solution. Compositional uniformity in the alloy product is enabled by use of the sandwich structure of uranium sheets and niobium powder.

This patent describes alloys such as U-6Nb prepared by forming a stacked sandwich array of uranium sheets and niobium powder disposed in layers between the sheets, heating the array in a vacuum induction melting furnace to a temperature such as to melt the uranium, holding the resulting mixture at a temperature above the melting point of uranium until the niobium dissolves in the uranium, and casting the uranium-niobium solution. Compositional uniformity in the alloy product is enabled by use of the sandwich structure of uranium sheets and niobium powder.

In early March 1994, eight highly enriched uranium (HEU) samples were collected from materials stored at the Ulba Metallurgical Plant in Oskamen (Ust Kamenogorsk), Kazakhstan. While at the plant site, portions of four samples were dissolved and analyzed by mass spectrograph at the Ulba analytical laboratory by Ulba analysts. Three of these mass spectrograph solutions and the eight HEU samples were subsequently delivered to the Y-12 Plant for complete chemical and isotopic analyses. Chemical forms of the eight samples were uranium metal chips, U0{sub 2} powder, uranium/beryllium oxide powder, and uranium/beryllium alloy rods. All were declared by the Ulba plant to have a uranium assay of {approximately}90 wt % {sup 235}U. The uranium/beryllium powder and alloy samples were also declared to range from about 8 to 28 wt % uranium. The chemical and uranium isotopic analyses done at the Y-12 Plant confirm the Ulba plant declarations. All samples appear to have been enriched using some reprocessed uranium, probably from recovery of uranium from plutonium production reactors. As a result, all samples contain some {sup 236}U and {sup 232}U and have small but measurable quantities of plutonium. This plutonium could be the result of either contamination carried over from the enrichment process or cross-contamination from weapons material. It is not the result of direct reactor exposure. Neither the {sup 232}U nor the plutonium concentrations are sufficiently high to provide a significant industrial health hazard. Both are well within established or proposed acceptance criteria for storage at Y-12. The trace metal analyses showed that, with the exception of beryllium, there are no trace metals in any of these HEU samples that pose a significant health hazard.

One of the most challenging issues facing the Department of Energy`s Office of Environmental Management is the cleanup of the three gaseous diffusion plants. In October 1992, Congress passed the Energy Policy Act of 1992 and established the Uranium Enrichment Decontamination and Decommissioning Fund to accomplish this task. This mission is being undertaken in an environmentally and financially responsible way by: devising cost-effective technical solutions; producing realistic life-cycle cost estimates, based on practical assumptions and thorough analysis; generating coherent long-term plans which are based on risk assessments, land use, and input from stakeholders; and, showing near-term progress in the cleanup of the gaseous diffusion facilities at Oak Ridge.

Depleted uranium hexafluoride (UF{sub 6}) has been produced in the United States since the 1940s as part of both the military program and the civilian nuclear energy program. The U.S. Department of Energy (DOE) is the agency responsible for managing most of the depleted UF{sub 6} that has been produced in the United States. The total quantity of depleted UF{sub 6} that DOE has to or will have to manage is approximately 700,000 Mg. Studies have been conducted to evaluate the various alternatives for managing this material. This paper evaluates and summarizes the alternative of disposal as low-level waste (LLW). Results of the analysis indicate that UF{sub 6} needs to be converted to a more stable form, such as U{sub 3}O{sub 8}, before disposal as LLW. Estimates of the environmental impacts of disposal in a dry environment are within the currently applicable standards and regulations. Of the currently operating LLW disposal facilities, available information indicates that either of two DOE facilities--the Hanford Site or the Nevada Test Site--or a commercial facility--Envirocare of Utah--would be able to dispose of up to the entire DOE inventory of depleted UF{sub 6}.

Based on official information received from 40 countries, Uranium 2007 provides a comprehensive review of world uranium supply and demand as of 1st January 2007, as well as data on global uranium exploration, resources, production and reactor-related requirements. It provides substantive new information from major uranium production centres in Africa, Australia, Central Asia, Eastern Europe and North America. Projections of nuclear generating capacity and reactor-related uranium requirements through 2030 are also featured, along with an analysis of long-term uranium supply and demand issues. It finds that with rising demand and declining inventories, uranium prices have increased dramatically in recent years. As a result, the uranium industry is undergoing a significant revival, bringing to an end a period of over 20 years of underinvestment.

An in situ recovery of uranium operation involves circulating reactive fluids through an underground uranium deposit. These fluids contain chemicals that dissolve the uranium ore. Uranium is recovered from the fluids after they are pumped back to the surface. Chemicals used to accomplish this include complexing agents that are organic, readily degradable, and/or have a predictable lifetime in an aquifer. Efficiency is increased through development of organic agents targeted to complexing tetravalent uranium rather than hexavalent uranium. The operation provides for in situ immobilization of some oxy-anion pollutants under oxidizing conditions as well as reducing conditions. The operation also artificially reestablishes reducing conditions on the aquifer after uranium recovery is completed. With the ability to have the impacted aquifer reliably remediated, the uranium recovery operation can be considered inherently safe.

that Geobacter species can effectively remove uranium from contaminated groundwater by reducing soluble U emplaced in flow- through columns of uranium-contaminated sediments readily removed U(VI) from the groundwater, and 87% of the uranium that had been removed was recovered from the electrode surface after

Alloys of uranium and tungsten and a method for making the alloys. The amount of tungsten present in the alloys is from about 4 wt % to about 35 wt %. Tungsten particles are dispersed throughout the uranium and a small amount of tungsten is dissolved in the uranium.

Alloys of uranium and tungsten and a method for making the alloys. The amount of tungsten present in the alloys is from about 4 wt % to about 35 wt %. Tungsten particles are dispersed throughout the uranium and a small amount of tungsten is dissolved in the uranium.

URANIUM MILL TAILINGS RADON FLUX CALCULATIONS PIĂ?ON RIDGE PROJECT MONTROSE COUNTY, COLORADO Inc. (Golder) was commissioned by EFRC to evaluate the operations of the uranium mill tailings storage in this report were conducted using the WISE Uranium Mill Tailings Radon Flux Calculator, as updated on November

The major health hazard from uranium mill tailings is presumed to be respiratory cancer resulting from the inhalation of radon daughter products. A review of studies on inhalation of radon and its daughters indicates that the hazard from the tailings is extremely small. If the assumptions used in the studies are correct, one or two people per year in the US may develop cancer as a result of radon exhaled from all the Uranium Mill Tailings Remedial Action Program sites. The remedial action should reduce the hazard from the tailings by a factor of about 100.

Food chain transfer through the soil-vegetation-small mammal food chain was measured by concentration ratios (CRs) for uranium, {sup 226}Ra, {sup 210}Pb, and {sup 210}Po at three sites near the Key Lake uranium mill in northern Saskatchewan. Plant/soil CRs, animal carcass/GI tract CRs, and animal/soil CRs were depressed at sites impacted by mill and tailings dusts relative to a nearby control site. Thus, radionuclides associated with large particulates in tailings and/or ore dusts may be less bioavailable to terrestrial plants and animals than natural sources of radioactive dust. These results show that reliance on default food chain transfer parameters, obtained from uncontaminated terrestrial ecosystems, may overpredict impacts at uranium mine and mill sites. Given the omnivorous diet of small mammals and birds, animal/soil CRs are recommended as the most cost-effective and robust means of predicting animal concentrations from environmental monitoring data at uranium mill facilities.

Solvent extraction has been applied to uranium-concentrate production in Brazil. At the first plant, uranium minerals associated with Zr and Mo were acid leached. Extraction was carried out by a mixture of Alamine 336 and Alamine 304, followed by selective Zr, U, and Mo stripping. At the currently operating facilities, a single U mineral is processed by acid heap leaching. Uranium is extracted with Alamine 336 and stripped with NaCl solution. As all water is recycled, chloride contents in the liquor have increased, causing detrimental effects to the extraction process. The current plant operating conditions and the improvements arisen from the research developed to solve these problems are presented. (authors)

Engineering disadvantages of unalloyed uranium include relatively low strength, low ductility, and poor oxidation and corrosion resistance. As-cast uranium typically exhibits very large grains that cause nonuniform deformation and low tensile ductility. Uranium is often alloyed to improve its corrosion resistance and mechanical properties. Titanium is most commonly used to increase strength; niobium and molybdenum, to increase oxidation and corrosion resistance; and vanadium, to refine alpha grain size in castings. Under equilibrium conditions these elements are extensively soluble in the high-temperature gamma phase, slightly soluble in the intermediate temperature beta phase, and essentially insoluble in the low-temperature alpha phase. Uranium alloys are vacuum solution heat treated in the gamma range to dissolve the alloying elements and remove hydrogen. The subsequent microstructures and properties are determined by the cooling rate from the solution treatment temperature. Oxidation and corrosion resistance increases with increasing the amount of alloy in solid solution. As a result, alloys such as U-6%Nb and U-10%Mo are often used in applications requiring good corrosion resistance.

Electron backscatter diffraction (EBSD) was used to examine the microstructures of unalloyed uranium, U-6Nb, U-10Mo, and U-0.75Ti. For unalloyed uranium, we used EBSD to examine the effects of various processes on microstructures including casting, rolling and forming, recrystallization, welding, and quasi-static and shock deformation. For U-6Nb we used EBSD to examine the microstructural evolution during shape memory loading. EBSD was used to study chemical homogenization in U-10Mo, and for U-0.75Ti, we used EBSD to study the microstructure and texture evolution during thermal cycling and deformation. The studied uranium alloys have significant microstructural and chemical differences and each of these alloys presents unique preparation challenges. Each of the alloys is prepared by a sequence of mechanical grinding and polishing followed by electropolishing with subtle differences between the alloys. U-6Nb and U-0.75Ti both have martensitic microstructures and both require special care in order to avoid mechanical polishing artifacts. Unalloyed uranium has a tendency to rapidly oxidize when exposed to air and a two-step electropolish is employed, the first step to remove the damaged surface layer resulting from the mechanical preparation and the second step to passivate the surface. All of the alloying additions provide a level of surface passivation and different one and two step electropolishes are employed to create good EBSD surfaces. Because of its low symmetry crystal structure, uranium exhibits complex deformation behavior including operation of multiple deformation twinning modes. EBSD was used to observe and quantify twinning contributions to deformation and to examine the fracture behavior. Figure 1 shows a cross section of two mating fracture surfaces in cast uranium showing the propensity of deformation twinning and intergranular fracture largely between dissimilarly oriented grains. Deformation of U-6Nb in the shape memory regime occurs by the motion of twin boundaries formed during the martensitic transformation. Deformation actually results in a coarsening of the microstructure making EBSD more practical following a limited amount of strain. Figure 2 shows the microstructure resulting from 6% compression. Casting of U-10Mo results in considerable chemical segregation as is apparent in Figure 2a. The segregation subsists through rolling and heat treatment processes as shown in Figure 2b. EBSD was used to study the effects of homogenization time and temperature on chemical heterogeneity. It was found that times and temperatures that result in a chemically homogeneous microstructure also result in a significant increase in grain size. U-0.75Ti forms an acicular martinsite as shown in Figure 4. This microstructure prevails through cycling into the higher temperature solid uranium phases.

A Depleted Uranium Silicate Container Backfill System (DUSCOBS) is proposed that would use small, isotopically-depleted uranium silicate glass beads as a backfill material inside repository waste packages containing spent nuclear fuel (SNF). The uranium silicate glass beads would fill the void space inside the package including the coolant channels inside SNF assemblies. Based on preliminary analysis, the following benefits have been identified. DUSCOBS improves repository waste package performance by three mechanisms. First, it reduces the radionuclide releases from SNF when water enters the waste package by creating a local uranium silicate saturated groundwater environment that suppresses (a) the dissolution and/or transformation of uranium dioxide fuel pellets and, hence, (b) the release of radionuclides incorporated into the SNF pellets. Second, the potential for long-term nuclear criticality is reduced by isotopic exchange of enriched uranium in SNF with the depleted uranium (DU) in the glass. Third, the backfill reduces radiation interactions between SNF and the local environment (package and local geology) and thus reduces generation of hydrogen, acids, and other chemicals that degrade the waste package system. Finally, DUSCOBS provides a potential method to dispose of significant quantities of excess DU from uranium enrichment plants at potential economic savings. DUSCOBS is a new concept. Consequently, the concept has not been optimized or demonstrated in laboratory experiments.

This invention is a simple and effective method for removing uranium from aqueous HF solutions containing trace quantities of the same. The method comprises contacting the solution with particulate calcium fluoride to form uranium-bearing particulates, permitting the particulates to settle, and separting the solution from the settled particulates. The CaF.sub.2 is selected to have a nitrogen surface area in a selected range and is employed in an amount providing a calcium fluoride/uranium weight ratio in a selected range. As applied to dilute HF solutions containing 120 ppm uranium, the method removes at least 92% of the uranium, without introducing contaminants to the product solution.

Alloys such as U-6Nb are prepared by forming a stacked sandwich array of uraniun sheets and niobium powder disposed in layers between the sheets, heating the array in a vacuum induction melting furnace to a temperature such as to melt the uranium, holding the resulting mixture at a temperature above the melting point of uranium until the niobium dissolves in the uranium, and casting the uranium-niobium solution. Compositional uniformity in the alloy product is enabled by use of the sandwich structure of uranium sheets and niobium powder.

Published every other year, Uranium Resources, Production, and Demand, or the "Red Book" as it is commonly known, is jointly prepared by the OECD Nuclear Energy Agency and the International Atomic Energy Agency. It is the recognised world reference on uranium and is based on official information received from 43 countries. It presents the results of a thorough review of world uranium supplies and demand and provides a statistical profile of the world uranium industry in the areas of exploration, resource estimates, production and reactor-related requirements. It provides substantial new information from all major uranium production centres in Africa, Australia, Central Asia, Eastern Europe and North America. Long-term projections of nuclear generating capacity and reactor-related uranium requirements are provided as well as a discussion of long-term uranium supply and demand issues. This edition focuses on recent price and production increases that could signal major changes in the industry.

Published every other year, Uranium Resources, Production, and Demand, or the "Red Book" as it is commonly known, is jointly prepared by the OECD Nuclear Energy Agency and the International Atomic Energy Agency. It is the recognised world reference on uranium and is based on official information received from 43 countries. This 21st edition presents the results of a thorough review of world uranium supplies and demand as of 1st January 2005 and provides a statistical profile of the world uranium industry in the areas of exploration, resource estimates, production and reactor-related requirements. It provides substantial new information from all major uranium production centres in Africa, Australia, Central Asia, Eastern Europe and North America. Projections of nuclear generating capacity and reactor-related uranium requirements through 2025 are provided as well as a discussion of long-term uranium supply and demand issues. This edition focuses on recent price and production increases that could signal major c...

This report evaluates nuclear criticality safety for the High-Enriched Uranium storage area of the X-345 building of the Portsmouth Gaseous Diffusion Plant. The effects of loss of moderation or mass control are examined for storage units in or out of the storage receptacles. Recommendations are made for decreasing criticality hazards under some conditions of storage or handling considered to be hazardous.

1.1 This specification covers nuclear grade uranium hexafluoride (UF6) that either has been processed through an enrichment plant, or has been produced by the blending of Highly Enriched Uranium with other uranium to obtain uranium of any 235U concentration below 5 % and that is intended for fuel fabrication. The objectives of this specification are twofold: (1) To define the impurity and uranium isotope limits for Enriched Commercial Grade UF6 so that, with respect to fuel design and manufacture, it is essentially equivalent to enriched uranium made from natural UF6; and (2) To define limits for Enriched Reprocessed UF6 to be expected if Reprocessed UF6 is to be enriched without dilution with Commercial Natural UF6. For such UF6, special provisions, not defined herein, may be needed to ensure fuel performance and to protect the work force, process equipment, and the environment. 1.2 This specification is intended to provide the nuclear industry with a standard for enriched UF6 that is to be used in the pro...

During the National Uranium Resource Evaluation (NURE) program, conducted for the US Department of Energy (DOE) by Bendix Field Engineering Corporation (BFEC), radiometric and geochemical surveys and geologic investigations detected anomalies indicative of possible uranium enrichment. Data from the Aerial Radiometric and Magnetic Survey (ARMS) and the Hydrogeochemical and Stream-Sediment Reconnaissance (HSSR), both of which were conducted on a national scale, yielded numerous anomalies that may signal areas favorable for the occurrence of uranium deposits. Results from geologic evaluations of individual 1/sup 0/ x 2/sup 0/ quadrangles for the NURE program also yielded anomalies, which could not be adequately checked during scheduled field work. Included in this volume are individual reports of field investigations for the following six areas which were shown on the basis of ARMS, HSSR, and (or) geologic data to be anomalous: (1) Hylas zone and northern Richmond basin, Virginia; (2) Sischu Creek area, Alaska; (3) Goodman-Dunbar area, Wisconsin; (4) McCaslin syncline, Wisconsin; (5) Mt. Withington Cauldron, Socorro County, New Mexico; (6) Lake Tecopa, Inyo County, California. Field checks were conducted in each case to verify an indicated anomalous condition and to determine the nature of materials causing the anomaly. The ultimate objective of work is to determine whether favorable conditions exist for the occurrence of uranium deposits in areas that either had not been previously evaluated or were evaluated before data from recent surveys were available. Most field checks were of short duration (2 to 5 days). The work was done by various investigators using different procedures, which accounts for variations in format in their reports. All papers have been abstracted and indexed.

An accurate outlook on long-term uranium resources is critical in forecasting uranium costresource relationships, and for energy policy planning as regards the development and deployment of nuclear fuel cycle alternatives. ...

Throughout the United States Department of Energy (DOE) complex, there are a number of streams of low enriched uranium (LEU) that contain various trace contaminants. These surplus nuclear materials require processing in order to meet commercial fuel cycle specifications. To date, they have not been designated as waste for disposal at the DOE's Nevada Test Site (NTS). Currently, with no commercial outlet available, the DOE is evaluating treatment and disposal as the ultimate disposition path for these materials. This paper will describe an innovative program that will provide a solution to DOE that will allow disposition of these materials at a cost that will be competitive with treatment and disposal at the NTS, while at the same time recycling the material to recover a valuable energy resource (yellowcake) for reintroduction into the commercial nuclear fuel cycle. International Uranium (USA) Corporation (IUSA) and Nuclear Fuel Services, Inc. (NFS) have entered into a commercial relationship to pursue the development of this program. The program involves the design of a process and construction of a plant at NFS' site in Erwin, Tennessee, for the blending of contaminated LEU with depleted uranium (DU) to produce a uranium source material ore (USM Ore{trademark}). The USM Ore{trademark} will then be further processed at IUC's White Mesa Mill, located near Blanding, Utah, to produce conventional yellowcake, which can be delivered to conversion facilities, in the same manner as yellowcake that is produced from natural ores or other alternate feed materials. The primary source of feed for the business will be the significant sources of trace contaminated materials within the DOE complex. NFS has developed a dry blending process (DRYSM Process) to blend the surplus LEU material with DU at its Part 70 licensed facility, to produce USM Ore{trademark} with a U235 content within the range of U235 concentrations for source material. By reducing the U235 content to source material levels in this manner, the material will be suitable for processing at a conventional uranium mill under its existing Part 40 license to remove contaminants and enable the product to re-enter the commercial fuel cycle. The tailings from processing the USM Ore{trademark} at the mill will be permanently disposed of in the mill's tailings impoundment as 11e.(2) byproduct material. Blending LEU with DU to make a uranium source material ore that can be returned to the nuclear fuel cycle for processing to produce yellowcake, has never been accomplished before. This program will allow DOE to disposition its surplus LEU and DU in a cost effective manner, and at the same time provide for the recovery of valuable energy resources that would be lost through processing and disposal of the materials. This paper will discuss the nature of the surplus LEU and DU materials, the manner in which the LEU will be blended with DU to form a uranium source material ore, and the legal means by which this blending can be accomplished at a facility licensed under 10 CFR Part 70 to produce ore that can be processed at a conventional uranium mill licensed under 10 CFR Part 40.

The Department of Energy (DOE), Office of Environmental Restoration and Waste Management, has chartered a study to evaluate alternative management strategies for depleted uranium (DU) currently stored throughout the DOE complex. Historically, DU has been maintained as a strategic resource because of uses for DU metal and potential uses for further enrichment or for uranium oxide as breeder reactor blanket fuel. This study has focused on evaluating the disposal options for DU if it were considered a waste. This report is in no way declaring these DU reserves a ``waste,`` but is intended to provide baseline data for comparison with other management options for use of DU. To PICS considered in this report include: Retrievable disposal; permanent disposal; health hazards; radiation toxicity and chemical toxicity.

. 2 to 25 ppm (Katz 1951). From thermal equilibrium calculations on the earth's core, mantle, and crust, and through actual analysis of samples, uranium was found to be concentrated in the earth's crust. According to modern geological thought..., as the uniformly molten earth cooled, its matter became separated into one vapor phase and three concentric condensed phases: the siderosphere (the earth's core, probably primarily molten iron), the chalcosphere forming the intermediate shell (the mantle...

An experimental study was conducted to determine the practicality of various unit operations for fluorination of uranium dioxide. The objective was to prepare ammonium uranium fluoride double salts from uranium dioxide and ...

A variety of analytical techniques is available for evaluating uranium in excreta and tissues at levels appropriate for occupational exposure control and evaluation. A few (fluorometry, kinetic phosphorescence analysis, {alpha}-particle spectrometry, neutron irradiation techniques, and inductively-coupled plasma mass spectrometry) have also been demonstrated as capable of determining uranium in these materials at levels comparable to those which occur naturally. Sample preparation requirements and isotopic sensitivities vary widely among these techniques and should be considered carefully when choosing a method. This report discusses analytical techniques used for evaluating uranium in biological matrices (primarily urine) and limits of detection reported in the literature. No cost comparison is attempted, although references are cited which address cost. Techniques discussed include: {alpha}-particle spectrometry; liquid scintillation spectrometry, fluorometry, phosphorometry, neutron activation analysis, fission-track counting, UV-visible absorption spectrophotometry, resonance ionization mass spectrometry, and inductively-coupled plasma mass spectrometry. A summary table of reported limits of detection and of the more important experimental conditions associated with these reported limits is also provided.

, we are currently investigating the coordina- tion chemistry of uranium metal centers with classicalUranium Tris-aryloxide Derivatives Supported by Triazacyclononane: Engendering a Reactive Uranium, and Karsten Meyer* Contribution from the Department of Chemistry and Biochemistry, UniVersity of California

Disclosed are methods, materials and systems that can be used to determine qualitatively or quantitatively the level of uranium contamination in water samples. Beneficially, disclosed systems are relatively simple and cost-effective. For example, disclosed systems can be utilized by consumers having little or no training in chemical analysis techniques. Methods generally include a concentration step and a complexation step. Uranium concentration can be carried out according to an extraction chromatographic process and complexation can chemically bind uranium with a detectable substance such that the formed substance is visually detectable. Methods can detect uranium contamination down to levels even below the MCL as established by the EPA.

Statistical Data of the Uranium Industry is a compendium of information relating to US uranium reserves and potential resources and to exploration, mining, milling, and other activities of the uranium industry through 1981. The statistics are based primarily on data provided voluntarily by the uranium exploration, mining, and milling companies. The compendium has been published annually since 1968 and reflects the basic programs of the Grand Junction Area Office (GJAO) of the US Department of Energy. The production, reserves, and drilling information is reported in a manner which avoids disclosure of proprietary information.

Abstract: Depleted uranium (DU) is an emerging environmental pollutant that is introduced into the environment primarily by military activity. While depleted uranium is less radioactive than natural uranium, it still retains all the chemical toxicity associated with the original element. In large doses the kidney is the target organ for the acute chemical toxicity of this metal, producing potentially lethal tubular necrosis. In contrast, chronic low dose exposure to depleted uranium may not produce a clear and defined set of symptoms. Chronic low-dose, or subacute, exposure to depleted uranium alters the appearance of milestones in developing organisms. Adult animals that were exposed to depleted uranium during development display persistent alterations in behavior, even after cessation of depleted uranium exposure. Adult animals exposed to depleted uranium demonstrate altered behaviors and a variety of alterations to brain chemistry. Despite its reduced level of radioactivity evidence continues to accumulate that depleted uranium, if ingested, may pose a radiologic hazard. The current state of knowledge concerning DU is discussed.

Alloys of tungsten and uranium and a method for making the alloys. The amount of tungsten present in the alloys is from about 55 vol % to about 85 vol %. A porous preform is made by sintering consolidated tungsten powder. The preform is impregnated with molten uranium such that (1) uranium fills the pores of the preform to form uranium in a tungsten matrix or (2) uranium dissolves portions of the preform to form a continuous uranium phase containing tungsten particles.

The B Plant facility, is located in the 200 East Area at the Hanford Site in south-central Washington State. It consists of two major operating areas: the B Plant Canyon Building, and the Waste Encapsulation and Storage Facility (WESF). The B Plant was originally designed to chemically process spent nuclear fuels. After this initial mission was completed, the plant was modified to provide for the separation of strontium and cesium, individually, from the fission productwaste stream following plutonium and uranium recovery from irradiated reactor fuels in the Plutonium-Uranium Extraction Plant (PUREX). The recovered, purified, and concentrated strontium and cesium solutions were then transferred to the WESF for conversion to solid compounds, encapsulation, and interim storage. After strontium and cesium removal, the remaining waste was transferred from B Plant to tank farms. B Plantis an operating facility that is required to ensure safe storage And management of the WESF cesium and strontium capsules, as well as a substantial radiological inventory remaining in the plant from previous campaigns. There are currently no production activities at B Plant, but several operating systems are required to accomplish the current B Plant mission.B Plant receives and stores various chemicals from commercial suppliers for treatment of low-level waste generated at WESF and B Plant, generation of demineralized water, and conditioning of water used in heating, ventilation, and air conditioning units. This report describes the maintenance of B Plant, including personnel training and schedules.

The Rocky Flats Plant (RFP) is a Department of Energy (DOE) facility where plutonium and uranium components were manufactured for nuclear weapons. During plant operations radioactivity was inadvertently released into the environment. This study was initiated to characterize the uranium present in surface-waters at RFP. Three drainage basins and natural ephemeral streams transverse RFP. The Woman Creek drainage basin traverses and drains the southern portion of the site. The Rock Creek drainage basin drains the northwestern portion of the plant complex. The Walnut Creek drainage basin traverses the western, northern, and northeastern portions of the RFP site. Dams, detention ponds, diversion structures, and ditches have been constructed at RFP to control the release of plant discharges and surface (storm water) runoff. The ponds located downstream of the plant complex on North Walnut Creek are designated A-1 through A-4. Ponds on South Walnut Creek are designated B-1 through B-5. The ponds in the Woman Creek drainage basin are designated C-1 and C-2. Water samples were collected from each pond and the uranium was characterized by TIMS measurement techniques.

In 1965, the Oak Ridge Gaseous Diffusion Plant (ORGDP), now the K-25 Site, conducted a series of tests in which bare cylinders of uranium hexafluoride (UF{sub 6}) were exposed to engulfing oil fires for the US Atomic Energy Commission (AEC), now the US Department of Energy (DOE). The tests are described and the results, conclusions, and observations are presented. Two each of the following types of cylinders were tested: 3.5-in.-diam {times} 7.5-in.-long cylinders of Monel (Harshaw), 5.0-in.-diam {times} 30-in.-long cylinders of Monel, and 8-in.-diam {times} 48-in.-long cylinders of nickel. The cylinders were filled approximately to the standard UF{sub 6} fill limits of 5, 55, and 250 lb, respectively, with a U-235 content of 0.22%. The 5-in.- and 8-in.-diam cylinders were tested individually with and without their metal valve covers. For the 3.5-in.-diam Harshaw cylinders and the 5.0-in.-diam cylinder without a valve cover the valves failed and UF{sub 6} was released. The remaining cylinders ruptured explosively in time intervals ranging from about 8.5 to 11 min.

In 1965, the Oak Ridge Gaseous Diffusion Plant (ORGDP), now the K-25 Site, conducted a series of tests in which bare cylinders of uranium hexafluoride (UF{sub 6}) were exposed to engulfing oil fires for the US Atomic Energy Commission (AEC), now the US Department of Energy (DOE). The tests are described and the results, conclusions, and observations are presented. Two each of the following types of cylinders were tested: 3.5-in.-diam {times} 7.5-in.-long cylinders of Monel (Harshaw), 5.0-in.-diam {times} x 30-in.-long cylinders of Monel, and 8-in.-diam {times} 48-in.-long cylinders of nickel. The cylinders were filled approximately to the standard UF{sub 6} fill limits of 5, 55, and 250 lb, respectively, with a U-235 content of 0.22%. The 5-in.- and 8-in.-diam cylinders were tested individually with and without their metal valve covers. For the 3.5-in.-diam Harshaw cylinders and the 5.0-in.-diam cylinder without a valve cover, the valves failed and UF{sub 6} was released. The remaining 6 cylinders ruptured explosively in time intervals ranging from about 8.5 to 11 min.

Radiochemical analysis was performed on samples of soil and eight species of common vegetation growing on the Anaconda uranium mill site, located in New Mexico. The concentrations of the long-lived radionuclides U-238, U-234, Th-230, Ra-226, and Pb-210 in these forage plants were determined. The sampling procedures and analytical laboratory methods used are described. The highest radionuclide concentration found in a forage species was 130 pCi of Ra-226 per gram dry weight for grass growing on the main tailings pile at Anaconda, where the surface soil activity of Ra-226 was 236 pCi/g. A comparison of shoots activity with that of roots and soil was used to determine a distribution index and uptake coefficient for each species. The distribution index, the ratio of root activity to shoot activity, ranged from 0.30 (Th-230) in galleta grass (Hilaria jamesii) to 38.0 (Ra-226) in Indian ricegrass (Oryzopsis hymenoides). In nearly all instances, the roots contained higher radionuclide concentrations. The uptake coefficient, the ratio of vegetation activity to soil activity, ranged from 0.69 (U-238) in Indian ricegrass roots to 0.01 (U-238) in four-wing saltbush (Atriplex canescans) shoots. The range of radionuclide concentrations in plants growing on the Anaconda mill site is compared to that in vegetation from a control site 20 km away.

In 1994, the International Atomic Energy Agency (IAEA) initiated an ambitious program of worldwide field trials to evaluate the utility of environmental monitoring for safeguards. Part of this program involved two extensive United States field trials conducted at the large uranium enrichment facilities. The Paducah operation involves a large low-enriched uranium (LEU) gaseous diffusion plant while the Portsmouth facilities include a large gaseous diffusion plant that has produced both LEU and high-enriched uranium (HEU) as well as an LEU centrifuge facility. As a result of the Energy Policy Act of 1992, management of the uranium enrichment operations was assumed by the US Enrichment Corporation (USEC). The facilities are operated under contract by Martin Marietta Utility Services. Martin Marietta Energy Systems manages the environmental restoration and waste management programs at Portsmouth and Paducah for DOE. These field trials were conducted. Samples included swipes from inside and outside process buildings, vegetation and soil samples taken from locations up to 8 km from main sites, and hydrologic samples taken on the sites and at varying distances from the sites. Analytical results from bulk analysis were obtained using high abundance sensitivity thermal ionization mm spectrometers (TIMS). Uranium isotopics altered from the normal background percentages were found for all the sample types listed above, even on vegetation 5 km from one of the enrichment facilities. The results from these field trials demonstrate that dilution by natural background uranium does not remove from environmental samples the distinctive signatures that are characteristic of enrichment operations. Data from swipe samples taken within the enrichment facilities were particularly revealing. Particulate analysis of these swipes provided a detailed ``history`` of both facilities, including the assays of the end product and tails for both facilities.

This document describes the validation of KENO V.a using the 27-group ENDF/B-IV cross section library for highly enriched uranium and beryllium neutronic systems, and is in accordance with ANSI/ANS-8.1-1983(R1988) requirements for calculational methods. The validation has been performed on a Hewlett Packard 9000/Series 700 Workstation at the Oak Ridge Y-12 Plant Nuclear Criticality Safety Department using the Oak Ridge Y-12 Plant Nuclear Criticality Safety Software code package. Critical experiments from LA-2203, UCRL-4975, ORNL-2201, and ORNL/ENG-2 have been identified as having the constituents desired for this validation as well as sufficient experimental detail to allow accurate construction of KENO V.a calculational models. The results of these calculations establish the safety criteria to be employed in future calculational studies of these types of systems.

The Uranium Management Group (UMG) was established at the Department of Energy's (DOE's) Oak Ridge Operations in 1999 as a mechanism to expedite the de-inventory of surplus uranium from the Fernald Environmental Management Project site. This successful initial venture has broadened into providing uranium material de-inventory and consolidation support to the Hanford site as well as retrieving uranium materials that the Department had previously provided to universities under the loan/lease program. As of December 31, 2001, {approx} 4,300 metric tons of uranium (MTU) have been consolidated into a more cost effective interim storage location at the Portsmouth site near Piketon, OH. The UMG continues to uphold its corporate support mission by promoting the Nuclear Materials Stewardship Initiative (NMSI) and the twenty-five (25) action items of the Integrated Nuclear Materials Management Plan (1). Before additional consolidation efforts may commence to remove excess inventory from Environmental Management closure sites and universities, a Programmatic Environmental Assessment (PEA) must be completed. Two (2) noteworthy efforts currently being pursued involve the investigation of re-use opportunities for surplus uranium materials and the recovery of usable uranium from the shutdown Portsmouth cascade. In summary, the UMG is available as a DOE complex-wide technical resource to promote the responsible management of surplus uranium.

Contamination of soils, water, and sediments by radionuclides and toxic metals from uranium mill tailings, nuclear fuel manufacturing and nuclear weapons production is a major concern. Studies of the mechanisms of biotransformation of uranium and toxic metals under various microbial process conditions has resulted in the development of two treatment processes: (1) stabilization of uranium and toxic metals with reduction in waste volume and (2) removal and recovery of uranium and toxic metals from wastes and contaminated soils. Stabilization of uranium and toxic metals in wastes is accomplished by exploiting the unique metabolic capabilities of the anaerobic bacterium, Clostridium sp. The radionuclides and toxic metals are solubilized by the bacteria directly by enzymatic reductive dissolution, or indirectly due to the production of organic acid metabolites. The radionuclides and toxic metals released into solution are immobilized by enzymatic reductive precipitation, biosorption and redistribution with stable mineral phases in the waste. Non-hazardous bulk components of the waste volume. In the second process uranium and toxic metals are removed from wastes or contaminated soils by extracting with the complexing agent citric acid. The citric-acid extract is subjected to biodegradation to recover the toxic metals, followed by photochemical degradation of the uranium citrate complex which is recalcitrant to biodegradation. The toxic metals and uranium are recovered in separate fractions for recycling or for disposal. The use of combined chemical and microbiological treatment process is more efficient than present methods and should result in considerable savings in clean-up and disposal costs.

The supercritical, reactor core melting and nuclear fuel leaking accidents have troubled fission reactors for decades, and greatly limit their extensive applications. Now these troubles are still open. Here we first show a possible perfect reactor, Molten-Salt Depleted-Uranium Reactor which is no above accident trouble. We found this reactor could be realized in practical applications in terms of all of the scientific principle, principle of operation, technology, and engineering. Our results demonstrate how these reactors can possess and realize extraordinary excellent characteristics, no prompt critical, long-term safe and stable operation with negative feedback, closed uranium-plutonium cycle chain within the vessel, normal operation only with depleted-uranium, and depleted-uranium high burnup in reality, to realize with fission nuclear energy sufficiently satisfying humanity long-term energy resource needs, as well as thoroughly solve the challenges of nuclear criticality safety, uranium resource insuffic...

Nine species of polyhydroxyanthraquinone and two of polyhydroxynaphthoquinone were screened to determine which have the greatest ability to accumulate uranium. 1,2-Dihydroxyanthraquinone and 3-amino-1,2-dihydroxyanthraquinone have extremely high accumulation abilities. To improve the adsorbing characteristics of these compounds, the authors tried to immobilize these compounds by coupling with diazotized aminopolystyrene. The immobilized 1,2-dihydroxyanthraquinone has the most favorable features for uranium recovery; high selective adsorption ability to uranium, rapid adsorption rate, and applicability in both column and batch systems. This adsorbent can recover uranium almost quantitatively from natural seawater. Almost all uranium adsorbed is desorbed with a solution of 1 N HCl. Thus, immobilized 1,2-dihydroxyanthraquinone can be used repeatedly in the adsorption-desorption process.

A program is underway at the Lawrence Livermore National Laboratory (LLNL) to recycle scrap uranium metal. Currently, much of the material from forging and machining processes is considered radioactive waste and is disposed of by oxidation and encapsulation at significant cost. In the recycling process, uranium and uranium alloys in various forms will be processed by electron beam melting and continuously cast into ingots meeting applicable specifications for virgin material. Existing vacuum processing facilities at LLNL are in compliance with all current federal and state environmental, safety and health regulations for the electron beam melting and vaporization of uranium metal. One of these facilities has been retrofitted with an auxiliary electron beam gun system, water-cooled hearth, crucible and ingot puller to create an electron beam melt furnace. In this furnace, basic process R&D on uranium recycling will be performed with the goal of eventual transfer of this technology to a production facility.

The present invention is directed to a process for fabricating laminated composites of uranium or uranium alloys and at least one other metal or alloy. The laminated composites are fabricated by forming a casting of the molten uranium with the other metal or alloy which is selectively positioned in the casting and then hot-rolling the casting into a laminated plate in or around which the casting components are metallurgically bonded to one another to form the composite. The process of the present invention provides strong metallurgical bonds between the laminate components primarily since the bond disrupting surface oxides on the uranium or uranium alloy float to the surface of the casting to effectively remove the oxides from the bonding surfaces of the components.

A conceptual design study of very small 350 MWth Gas-cooled Fast Reactors with Helium coolant has been performed. In this study Modified CANDLE burn-up scheme was implemented to create small and long life fast reactors with natural Uranium as fuel cycle input. Such system can utilize natural Uranium resources efficiently without the necessity of enrichment plant or reprocessing plant. The core with metallic fuel based was subdivided into 10 regions with the same volume. The fresh Natural Uranium is initially put in region-1, after one cycle of 10 years of burn-up it is shifted to region-2 and the each region-1 is filled by fresh Natural Uranium fuel. This concept is basically applied to all axial regions. The reactor discharge burn-up is 31.8% HM. From the neutronic point of view, this design is in compliance with good performance.

A conceptual design study of Gas Cooled Fast Reactors with Modified CANDLE burn-up scheme has been performed. In this study, design GCFR with Helium coolant which can be continuously operated by supplying mixed Natural Uranium/Thorium without fuel enrichment plant or fuel reprocessing plant. The active reactor cores are divided into two region, Thorium fuel region and Uranium fuel region. Each fuel core regions are subdivided into ten parts (region-1 until region-10) with the same volume in the axial direction. The fresh Natural Uranium and Thorium is initially put in region-1, after one cycle of 10 years of burn-up it is shifted to region-2 and the each region-1 is filled by fresh natural Uranium/Thorium fuel. This concept is basically applied to all regions in both cores area, i.e. shifted the core of i{sup th} region into i+1 region after the end of 10 years burn-up cycle. For the next cycles, we will add only Natural Uranium and Thorium on each region-1. The calculation results show the reactivity reached by mixed Natural Uranium/Thorium with volume ratio is 4.7:1. This reactor can results power thermal 550 MWth. After reactor start-up the operation, furthermore reactor only needs Natural Uranium/Thorium supply for continue operation along 100 years.

Sources of potential releases of /sup 99/Tc to the environment are reviewed for the uranium fuel cycle considering two options: the recycle of spent uranium fuel and no fuel recycling. In the no recycle option, the only source of /sup 99/Tc release is an extremely small amount associated with airborne emissions from the processing of high-level wastes. With recycling, /sup 99/Tc releases are associated with the operation of reprocessing facilities, UF/sub 6/ conversion plants, uranium enrichment plants, fuel fabrication facilities, and low- and high-level waste processing and storage facilities. Among these, the most prominent /sup 99/Tc releases are from the liquid effluents of uranium enrichment facilities (0.22 Ci per reference reactor year). A review of parameters of importance for predicting the environmental behavior and fate of /sup 99/Tc indicates a substantial reduction from earlier estimates of the radiological significance of exposure pathways involving the ingestion of milk and meat. More important routes of exposure to /sup 99/Tc will probably be associated with drinking water and the consumption of aquatic organisms, garden vegetables, and eggs. For each parameter reviewed in this study, a range of values is recommended for radiological assessment calculations. Where obvious discrepancies exist between these ranges and the default values listed in USNRC Regulatory Guide 1.109, consideration for revision of the USNRC default values is recommended.

During the period 1947 through 1970, the Atomic Energy Commission (AEC) fostered the rapid development and expansion of the domestic uranium mining and milling industry by providing a market for uranium. Some thirty-two mills were constructed during that period to produce U/sub 3/O/sub 8/ concentrates for sale to the AEC. In addition, there were various pilot plants, concentrators, upgraders, heap leach, and solution mining facilities that operated during the period. The purpose of this report is to compile a short narrative history of the AEC's uranium concentrate procurement program and to describe briefly each of the operations that produced uranium for sale to the AEC. Contractual arrangements are described and data are given on quantities of U/sub 3/O/sub 8/ purchased and prices paid. Similar data are included for V/sub 2/O/sub 5/, where applicable. Mill and other plant operating data were also compiled from old AEC records. These latter data were provided by the companies, as a contractual requirement, during the period of operation under AEC contracts. Additionally, an effort was made to determine the present status of each facility by reference to other recently published reports. No sites were visited nor were the individual reports reviewed by the companies, many of which no longer exist. The authors relied almost entirely on published information for descriptions of facilities and milling processes utilized.

Some attempts were made to remove uranium that may be present in refining effluents, mine tailings by using new microorganisms isolated from uranium deposits and peculiar natural environments. To screen microorganisms isolated from uranium deposits and peculiar natural environments in North America and Japan for maximal accumulation of uranium, hundreds of microorganisms were examined. Some microorganisms can accumulate about 500 mg (4.2 mEq) of uranium per gram of Microbial cells within 1 h. The uranium accumulation capacity of the cells exceeds that of commercially available chelating agents (2-3 mEq/g adsorbent). We attempted to recover uranium from uranium refining waste water by using new microorganisms. As a result, these microbial cells can recover trace amounts of uranium from uranium waste water with high efficiency. These strains also have a high accumulating ability for thorium. Thus, these new microorganisms can be used as an adsorbing agent for the removal of nuclear elements may be present in metallurgical effluents, mine tailings and other waste sources.

Uranium Cluster Chemistry DOI: 10.1002/anie.200906605 Tetranuclear Uranium Clusters by Reductive in the coordination chemistry and small-molecule reactivity of uranium. Among the intriguing reactivity patterns of tetravalent uranium with 3,5-dimethylpyrazolate (Me2PzŔ ) led to forma- tion of an unprecedented homoleptic

The Urals Electrochemical Integrated Plant is one of the Russian Ministry of Atomic Energy`s nuclear material production sites participating in the US Department of Energy`s Material Protection, Control and Accounting (MPC&A) Program. The Urals Electrochemical Integrated Plant is Russia`s largest uranium enrichment facility and blends tons of high-enriched uranium into low enriched uranium each year as part of the US high-enriched uranium purchase. The Electrochemical Integrated Plant and six participating national laboratories are cooperating to implement a series of enhancements to the nuclear material protection, control, and accountability systems at the site This paper outlines the overall objectives of the MPC&A program at Urals Electrochemical Integrated Plant and the work completed as of the date of the presentation.

In situ bioremediation of uranium holds significant promise for effective stabilization of U(VI) from groundwater at reduced cost compared to conventional pump and treat. This promise is unlikely to be realized unless researchers and practitioners successfully predict and demonstrate the long-term effectiveness of uranium bioremediation protocols. Field research to date has focused on both proof of principle and a mechanistic level of understanding. Current practice typically involves an engineering approach using proprietary amendments that focuses mainly on monitoring U(VI) concentration for a limited time period. Given the complexity of uranium biogeochemistry and uranium secondary minerals, and the lack of documented case studies, a systematic monitoring approach using multiple performance indicators is needed. This document provides an overview of uranium bioremediation, summarizes design considerations, and identifies and prioritizes field performance indicators for the application of uranium bioremediation. The performance indicators provided as part of this document are based on current biogeochemical understanding of uranium and will enable practitioners to monitor the performance of their system and make a strong case to clients, regulators, and the public that the future performance of the system can be assured and changes in performance addressed as needed. The performance indicators established by this document and the information gained by using these indicators do add to the cost of uranium bioremediation. However, they are vital to the long-term success of the application of uranium bioremediation and provide a significant assurance that regulatory goals will be met. The document also emphasizes the need for systematic development of key information from bench scale tests and pilot scales tests prior to full-scale implementation.

In a study designed to provide an improved scientific basis for assessing possible health effects from inhaling depleted uranium (DU) aerosols, a series of DU penetrators was fired at an Abrams tank and a Bradley fighting vehicle. A robust sampling system was designed to collect aerosols in this difficult environment and continuously monitor the sampler flow rates. Aerosols collected were analyzed for uranium concentration and particle size distribution as a function of time. They were also analyzed for uranium oxide phases, particle morphology, and dissolution in vitro. The resulting data provide input useful in human health risk assessments.

An electrolytic process for making uranium from uranium oxide using Cl.sub.2 anode product from an electrolytic cell to react with UO.sub.2 to form uranium chlorides. The chlorides are used in low concentrations in a melt comprising fluorides and chlorides of potassium, sodium and barium in the electrolytic cell. The electrolysis produces Cl.sub.2 at the anode that reacts with UO.sub.2 in the feed reactor to form soluble UCl.sub.4, available for a continuous process in the electrolytic cell, rather than having insoluble UO.sub.2 fouling the cell.

To help assure that analysis data of known quality will be produced in support of demonstration programs at the Fuel Conditioning Facility at Argonne National Laboratory-West (Idaho Falls, ID), a special reference material has been prepared and characterized. Designated SRM A, the material consists of individual units of LiCl/KCl eutectic salt containing a nominal concentration of 2.5 wt. % enriched uranium. Analyses were performed at Argonne National Laboratory-East (Argonne, IL) to determine the uniformity of the material and to establish reference values for the uranium concentration and uranium isotopic composition. Ten units from a batch of approximately 190 units were analyzed by the mass spectrometric isotope dilution technique to determine their uranium concentration. These measurements provided a mean value of 2.5058 {+-} 0.0052 wt. % U, where the uncertainty includes estimated limits to both random and systematic errors that might have affected the measurements. Evidence was found of a small, apparently random, non-uniformity in uranium content of the individual SRM A units, which exhibits a standard deviation of 0.078% of the mean uranium concentration. Isotopic analysis of the uranium from three units, by means of thermal ionization mass spectrometry with a special, internal-standard procedure, indicated that the uranium isotopy is uniform among the pellets with a composition corresponding to 0.1115 {+-} 0.0006 wt. % {sup 234}U, 19.8336 {+-} 0.0059 wt. % {sup 235}U, 0.1337 {+-} 0.0006 wt. % {sup 236}U, and 79.9171 {+-} 0.0057 wt. % {sup 238}U.

Significant volumes of soil containing elevated levels of uranium exist in the eastern United States. The contamination resulted from the development of the nuclear industry in the United States requiring a large variety of uranium products. The contaminated soil poses a collection and disposal problem of a magnitude that justifies the development of decontamination methods. Consequently, the Department of Energy (DOE) Office of Technology Development formed the Uranium Soils Integrated Demonstration (USID) program to address the problem. The fundamental goal of the USID task group has been the selective extraction/leaching or removal of uranium from soil faster, cheaper, and safer than what can be done using current conventional technologies. The objective is to selectively remove uranium from soil without seriously degrading the soil`s physicochemical characteristics and without generating waste that is difficult to manage and/or dispose of. However, procedures developed for removing uranium from contaminated soil have involved harsh chemical treatments that affect the physicochemical properties of the soil. The questions are (1) are the changes in soil properties severe enough to destroy the soil`s capacity to support and sustain vegetation growth and survival? and (2) what amendments might be made to the leached soil to return it to a reasonable vegetation production capacity? This study examines the vegetation-support capacity of soil that had been chemically leached to remove uranium. The approach is to conduct short-term germination and phytotoxicity tests for evaluating soils after they are subjected to various leaching procedures followed by longer term pot studies on successfully leached soils that show the greatest capacity to support plant growth. This report details the results from germination and short-term phytotoxicity testing of soils that underwent a variety of leaching procedures at the bench scale at ORNL and at the pilot plant at Fernald.

This invention is a method for lowering the beta radiation hazards associated with the casting of uranium. The method reduces the beta radiation emitted from the as-cast surfaces of uranium ingots. The method also reduces the amount of beta radiation emitters retained on the interiors of the crucibles that have been used to melt the uranium charges and which have undergone cleaning in a remote handling facility. The lowering of the radioactivity is done by scavenging the beta emitters from the molten uranium with a molten mixture containing the fluorides of magnesium and calcium. The method provides a means of collection and disposal of the beta emitters in a manner that reduces radiation exposure to operating personnel in the work area where the ingots are cast and processed.

An amidoxime fibrous adsorbent is contacted with uranium-enriched seawater (10 ppm); about 10 mg uranium is loaded per 1 g dry fiber. Then the rate and yield of uranium desorption from the fiber are determined with various eluents. Acid solutions are superior to alkali carbonate solutions as eluents. With a 0.1 mol[center dot]L[sup [minus]1] HCl solution, desorption is completed in 2 hours regardless of the presence of uranium in the leaching solution up to 15 ppm ([approx]6 [times] 10[sup [minus]5]mol[center dot]L[sup [minus]1]). Serial operation of the adsorption-desorption cycle four times does not affect desorption efficiency, but the addition of heavy metal ions to the eluent at a level of 1.8 [times] 10[sup [minus]3]mol[center dot]L[sup [minus]1] significantly decreases desorption efficiency. 13 refs., 5 figs., 1 tab.

U.S. Department of Energy synthetic urine quality assurance standards from an inter-laboratory exercise in 2012. The separation apparatus was able to consistently separate uranium from the synthetic urine solutions with a consistent recovery between...

Sodium Fast Reactors are one of the three candidates of GEN-IV fast reactors. Fast reactors play an important role in saving uranium resources and reducing nuclear wastes. Conventional fast reactors rely on transuranic ...

In an effort to develop cost-efficient techniques for remediating uranium contaminated groundwater at DOE Uranium Mill Tailing Remedial Action (UMTRA) sites nationwide, Sandia National Laboratories (SNL) deployed a pilot scale research project at an UMTRA site in Durango, CO. Implementation included design, construction, and subsequent monitoring of an in situ passive reactive barrier to remove Uranium from the tailings pile effluent. A reactive subsurface barrier is produced by emplacing a reactant material (in this experiment - various forms of metallic iron) in the flow path of the contaminated groundwater. Conceptually the iron media reduces and/or adsorbs uranium in situ to acceptable regulatory levels. In addition, other metals such as Se, Mo, and As have been removed by the reductive/adsorptive process. The primary objective of the experiment was to eliminate the need for surface treatment of tailing pile effluent. Experimental design, and laboratory and field preliminary results are discussed with regard to other potential contaminated groundwater treatment applications.

In an effort to develop cost-efficient techniques for remediating uranium contaminated groundwater at DOE Uranium Mill Tailing Remedial Action (UMTRA) sites nationwide, Sandia National Laboratories (SNL) deployed a pilot scale research project at an UMTRA site in Durango, CO. Implementation included design, construction, and subsequent monitoring of an in situ passive reactive barrier to remove Uranium from the tailings pile effluent. A reactive subsurface barrier is produced by emplacing a reactant material (in this experiment various forms of metallic iron) in the flow path of the contaminated groundwater. Conceptually the iron media reduces and/or adsorbs uranium in situ to acceptable regulatory levels. In addition, other metals such as Se, Mo, and As have been removed by the reductive/adsorptive process. The primary objective of the experiment was to eliminate the need for surface treatment of tailing pile effluent. Experimental design, and laboratory and field results are discussed with regard to other potential contaminated groundwater treatment applications.

This invention is a method for lowering the beta radiation hazards associated with the casting of uranium. The method reduces the beta radiation emitted from the as-cast surfaces of uranium ingots. The method also reduces the amount of beta radiation emitters retained on the interiors of the crucibles that have been used to melt the uranium charges and which undergone cleaning in a remote handling facility. The lowering of the radioactivity is done by scavenging the beta emitters from the molten uranium with a molten mixture containing the fluorides of magnesium and calcium. The method provides a means of collection and disposal of the beta emitters in a manner that reduces radiation exposure to operating personnel in the work area where the ingots are cast and processed. 5 tabs.

A total of 61 critical configurations are reported for experiments involving various combinations of annular tanks into which enriched uranium solution was pumped. These experiments were performed at two widely separated times in the 1980s under two programs at the Rocky Flats Plant`s Critical Mass Laboratory. The uranyl nitrate solution contained about 370 g of uranium per liter, but this concentration varied a little over the duration of the studies. The uranium was enriched to about 93% [sup 235]U. All tanks were typical of sizes commonly found in nuclear production plants. They were about 2 m tall and ranged in diameter from 0.6 m to 1.5 m. Annular thicknesses and conditions of neutron reflection, moderation, and absorption were such that criticality would be achieved with these dimensions. Only 13 of the entire set of 74 experiments proved to be subcritical when tanks were completely filled with solution. Single tanks of several radial thicknesses were studied as well as small line arrays (1 x 2 and 1 x 3) of annular tanks. Many systems were reflected on four sides and the bottom by concrete, but none were reflected from above. Many experiments also contained materials within and outside the annular regions that contained strong neutron absorbers. One program had such a thick external moderator/absorber combination that no reflector was used at all.

. This monitoring is often multi-faceted and typically involves an air sampling and biological sampling regime. The regime depends on the potential for exposures, the materials and chemical compounds being used, and the facility history. Specifically... Y-12 led the early US uranium enrichment programs, it also pioneered early uranium bioassay.[8] Likewise, the 5 Savannah River Site (SRS) pioneered plutonium bioassay techniques.[9] From these programs, techniques were developed to detect...

Contamination of soils, water, and sediments by radionuclides and toxic metals from uranium mill tailings, nuclear fuel manufacturing and nuclear weapons production is a major concern. Studies of the mechanisms of biotransformation of uranium and toxic metals under various microbial process conditions has resulted in the development of two treatment processes: (i) stabilization of uranium and toxic metals with reduction in waste volume and (ii) removal and recovery of uranium and toxic metals from wastes and contaminated soils. Stabilization of uranium and toxic metals in wastes is accomplished by exploiting the unique metabolic capabilities of the anaerobic bacterium, Clostridium sp. The radionuclides and toxic metals are solubilized by the bacteria directly by enzymatic reductive dissolution, or indirectly due to the production of organic acid metabolites. The radionuclides and toxic metals released into solution are immobilized by enzymatic reductive precipitation, biosorption and redistribution with stable mineral phases in the waste. Non-hazardous bulk components of the waste such as Ca, Fe, K, Mg and Na released into solution are removed, thus reducing the waste volume. In the second process uranium and toxic metals are removed from wastes or contaminated soils by extracting with the complexing agent citric acid. The citric-acid extract is subjected to biodegradation to recover the toxic metals, followed by photochemical degradation of the uranium citrate complex which is recalcitrant to biodegradation. The toxic metals and uranium are recovered in separate fractions for recycling or for disposal. The use of combined chemical and microbiological treatment process is more efficient than present methods and should result in considerable savings in clean-up and disposal costs.

An apparatus and method for continuous production of liquid uranium alloys through the electrolytic reduction of uranium chlorides. The apparatus includes an electrochemical cell formed from an anode shaped to form an electrolyte reservoir, a cathode comprising a metal, such as iron, capable of forming a eutectic uranium alloy having a melting point less than the melting point of pure uranium, and molten electrolyte in the reservoir comprising a chlorine or fluorine containing salt and uranium chloride. The method of the invention produces an eutectic uranium alloy by creating an electrolyte reservoir defined by a container comprising an anode, placing an electrolyte in the reservoir, the electrolyte comprising a chlorine or fluorine containing salt and uranium chloride in molten form, positioning a cathode in the reservoir where the cathode comprises a metal capable of forming an uranium alloy having a melting point less than the melting point of pure uranium, and applying a current between the cathode and the anode.

The physicochemical characterization of uranium(VI) in seawater is described on the basis of species distribution calculations and experiments using polarography and spectrophotometry in artificial seawater at elevated uranium concentrations. Various dissolved uranium(VI) species are identified under different conditions of pH and carbonate concentration. Below pH 4, the hydrated uranyl ion is present in the free state (forming labile complexes). Above pH 4, a stepwise coordination of uranyl by the carbonate ion occurs. The monocarbonate complex is formed in the pH range 4-5, the bicarbonate uranyl complex between 5 and 6. Above pH 8, uranium is present predominately as the tricarbonate and to a smaller extent as a trihydroxide complex. There is satisfactory agreement between our experiments and the theoretically computed distribution of uranium(VI) in seawater based on published stability constants. The experiments done at higher concentrations are justified by theoretical distributions showing that there is no great difference in species distribution between the uranium at concentrations of 10/sup -4/ and /sup -8/ mol dm/sup -3/.

This report describes the Conversion and Blending Facility (CBF) which will have two missions: (1) convert surplus HEU materials to pure HEU UF{sub 6} and a (2) blend the pure HEU UF{sub 6} with diluent UF{sub 6} to produce LWR grade LEU-UF{sub 6}. The primary emphasis of this blending be to destroy the weapons capability of large, surplus stockpiles of HEU. The blended LEU product can only be made weapons capable again by the uranium enrichment process. The chemical and isotopic concentrations of the blended LEU product will be held within the specifications required for LWR fuel. The blended LEU product will be offered to the United States Enrichment Corporation (USEC) to be sold as feed material to the commercial nuclear industry.

This patent describes a method of removing niobium from a uranium-niobium alloy. It comprises dissolving the uranium-niobium alloy metal pieces in a first aqueous solution containing an acid selected from the group consisting of hydrochloric acid and sulfuric acid and fluoboric acid as a catalyst to provide a second aqueous solution, which includes uranium (U{sup +4}), acid radical ions, the acids insolubles including uranium oxides and niobium oxides; adding nitric acid to the insolubles to oxidize the niobium oxides to yield niobic acid and to complete the solubilization of any residual uranium; and separating the niobic acid from the nitric acid and solubilized uranium.

A computer program entitled URPE (Uranium Recovery Performance and Economics) has been developed for analysis of a plant recovering uranium from seawater. The conceptual system design consists of a floating oil-rig type platform, using seawater forced through hydrous titanium oxide. Uranium is recovered from the seawater by adsorption and eluted later. The equilibrium isotherm and the diffusion constant for the uranyl-HTO system, which are needed for bed performance calculations, have been calculated. The URPE program has been benchmarked against previous studies by ORNL and Exxon, and found to make comparable performance and economic estimates. The URPE code was then used to identify optimum bed operating conditions. Thin beds of small, thinly-coated particles are the preferred bed configuration, and actively pumped systems outperform current driven units. Based on URPE, the minimum expected costs of uranium recovered from seawater would be no lower than approx. 316 (1979$)/lb U/sub 3/O/sub 8/ for state-of-the-art adsorber material (capacity equal to 210 mg U/kg Ti), but might be reduced to the level of breakeven attractiveness of approx. 150 (1979$)/lb U/sub 3/U/sub 8/ if at least a four-fold increase in adsorption capacity could be achieved.

Elucidation of complex biogeochemical processes and their effects on speciation of U in the subsurface is critical for developing remediation strategies with an understanding of stability. We have developed static microcosms that are similar to bioreduction process studies in situ under laminar flow conditions or in sediment pores. Uranium L{sub 3}-edge X-ray absorption near-edge spectroscopy analysis with depth in the microcosms indicated that transformation of U{sup VI} to U{sup IV} occurred by at least two distinct processes. Extended X-ray absorption fine structure (EXAFS) analysis indicated that initial U{sup VI} species associated with C- and P-containing ligands were transformed to U{sup IV} in the form of uraninite and U associated with Fe-bound ligands. Microbial community analysis identified putative Fe{sup III} and sulfate reducers at two different depths in the microcosms. The slow reduction of U{sup VI} to U{sup IV} may contribute the stability of U{sup IV} within microcosms at 11 months after a decrease in bioreducing conditions due to limited electron donors.

The future viability of nuclear power will depend on the long-term availability of uranium. A two-form uranium supply model was used to estimate the date at which peak production will occur. The model assumes a constant ...

This study was conducted to test the ability of the Chemchek™ Kinetic Phosphorescence Analyzer Model KPA-11 with an auto-sampler to resolve the difference in phosphorescent decay rates of several different uranium isotopes, and therefore identify the uranium isotope ratios present in a sample. Kinetic phosphorescence analysis (KPA) is a technique that provides rapid, accurate, and precise determination of uranium concentration in aqueous solutions. Utilizing a pulsed-laser source to excite an aqueous solution of uranium, this technique measures the phosphorescent emission intensity over time to determine the phosphorescence decay profile. The phosphorescence intensity at the onset of decay is proportional to the uranium concentration in the sample. Calibration with uranium standards results in the accurate determination of actual concentration of the sample. Different isotopes of uranium, however, have unique properties which should result in different phosphorescence decay rates seen via KPA. Results show that a KPA is capable of resolving uranium isotopes.

This document summarizes environmental information which has been collected up to June 1983 at Savannah River Plant. Of particular interest is an updating of dose estimates from changes in methodology of calculation, lower cesium transport estimates from Steel Creek, and new sports fish consumption data for the Savannah River. The status of various permitting requirements are also discussed.

of Energy's Environmental Remediation Sciences Program. Q: How can uranium be removed or neutralized so in the contaminated subsurface and engineering the subsurface environment to stimulate nitrate removal and uraniumMicrobial Janitors: Enabling natural microbes to clean up uranium contamination Oak Ridge

", uranium is removed from the processed ore with sulfuric acid. Sodium chlorate is also addedAppendix IV. Risks Associated with Conventional Uranium Milling Operations Introduction Although uranium mill tailings are considered byproduct materials under the AEA and not TENORM, EPA's Science

A process is described for separating uranium values and transuranic values from fission products containing rare earth values when the values are contained together in a molten chloride salt electrolyte. A molten chloride salt electrolyte with a first ratio of plutonium chloride to uranium chloride is contacted with both a solid cathode and an anode having values of uranium and fission products including plutonium. A voltage is applied across the anode and cathode electrolytically to transfer uranium and plutonium from the anode to the electrolyte while uranium values in the electrolyte electrolytically deposit as uranium metal on the solid cathode in an amount equal to the uranium and plutonium transferred from the anode causing the electrolyte to have a second ratio of plutonium chloride to uranium chloride. Then the solid cathode with the uranium metal deposited thereon is removed and molten cadmium having uranium dissolved therein is brought into contact with the electrolyte resulting in chemical transfer of plutonium values from the electrolyte to the molten cadmium and transfer of uranium values from the molten cadmium to the electrolyte until the first ratio of plutonium chloride to uranium chloride is reestablished.

A process for separating uranium values and transuranic values from fission products containing rare earth values when the values are contained together in a molten chloride salt electrolyte. A molten chloride salt electrolyte with a first ratio of plutonium chloride to uranium chloride is contacted with both a solid cathode and an anode having values of uranium and fission products including plutonium. A voltage is applied across the anode and cathode electrolytically to transfer uranium and plutonium from the anode to the electrolyte while uranium values in the electrolyte electrolytically deposit as uranium metal on the solid cathode in an amount equal to the uranium and plutonium transferred from the anode causing the electrolyte to have a second ratio of plutonium chloride to uranium chloride. Then the solid cathode with the uranium metal deposited thereon is removed and molten cadmium having uranium dissolved therein is brought into contact with the electrolyte resulting in chemical transfer of plutonium values from the electrolyte to the molten cadmium and transfer of uranium values from the molten cadmium to the electrolyte until the first ratio of plutonium chloride to uranium chloride is reestablished.

This paper describes the validation in accordance with ANSI/ANS-8.1-1983(R1988) of KENO V.a using the 27-group ENDF/B-IV cross-section library for systems containing highly-enriched uranium, carbon, and hydrogen and for systems containing highly-enriched uranium and carbon with high carbon to uranium (C/U) atomic ratios. The validation has been performed for two separate computational platforms: an IBM 3090 mainframe and an HP 9000 Model 730 workstation, both using the Oak Ridge Y-12 Plant Nuclear Criticality Safety Software (NCSS) code package. Critical experiments performed at the Oak Ridge Critical Experiments Facility, in support of the Rover reactor program, and at the Pajarito site at Los Alamos National Laboratory were identified as having the constituents desired for this validation as well as sufficient experimental detail to allow accurate construction of KENO V.a calculational models. Calculated values of k{sub eff} for the Rover experiments, which contain uranium, carbon, and hydrogen, are between 1.0012 {+-} 0.0026 and 1.0245 {+-} 0.0023. Calculation of the Los Alamos experiments, which contain uranium and carbon at high C/U ratios, yields values of k{sub eff} between 0.9746 {+-} 0.0028 and 0.9983 {+-} 0.0027. Safety criteria can be established using this data for both types of systems.

the US Department of Energy is evaluating technologies for the storage, disposal, or re-use of depleted uranium hexafluoride (UF{sub 6}). This paper discusses the following options, and provides a technology assessment for each one: (1) conversion to UO{sub 2} for use as mixed oxide duel, (2) conversion to UO{sub 2} to make DUCRETE for a multi-purpose storage container, (3) conversion to depleted uranium metal for use as shielding, (4) conversion to uranium carbide for use as high-temperature gas-cooled reactor (HTGR) fuel. In addition, conversion to U{sub 3}O{sub 8} as an option for long-term storage is discussed.

A system life-cycle cost study was conducted of a preliminary design concept for a plasma reduction process for converting depleted uranium to uranium metal and anhydrous HF. The plasma-based process is expected to offer significant economic and environmental advantages over present technology. Depleted Uranium is currently stored in the form of solid UF{sub 6}, of which approximately 575,000 metric tons is stored at three locations in the U.S. The proposed system is preconceptual in nature, but includes all necessary processing equipment and facilities to perform the process. The study has identified total processing cost of approximately $3.00/kg of UF{sub 6} processed. Based on the results of this study, the development of a laboratory-scale system (1 kg/h throughput of UF6) is warranted. Further scaling of the process to pilot scale will be determined after laboratory testing is complete.

Highly pure uranium hexafluoride is made from uranium oxide and fluorine. The uranium oxide, which includes UO.sub.3, UO.sub.2, U.sub.3 O.sub.8 and mixtures thereof, is introduced together with a small amount of a fluorine-reactive substance, selected from alkali chlorides, silicon dioxide, silicic acid, ferric oxide, and bromine, into a constant volume reaction zone. Sufficient fluorine is charged into the zone at a temperature below approximately 0.degree. C. to provide an initial pressure of at least approximately 600 lbs/sq. in. at the ambient atmospheric temperature. The temperature is then allowed to rise in the reaction zone until reaction occurs.

The active well coincidence counter (AWCC) was calibrated at the Chalk River Nuclear Laboratories (CRNL) for the assay of 93%-enriched fuel materials in three categories: (1) uranium-aluminum billets, (2) uranium-aluminum fuel elements, and (3) uranium metal pieces. The AWCC was a standard instrument supplied to the International Atomic Energy Agency under the International Safeguards Project Office Task A.51. Excellent agreement was obtained between the CRNL measurements and previous Los Alamos National Laboratory measurements on similar mockup fuel material. Calibration curves were obtained for each sample category. 2 refs., 8 figs., 15 tabs.

With the publication of a Request for Recommendations and Advance Notice of Intent in the November 10, 1994 Federal Register, the Department of Energy initiated a program to assess alternative strategies for the long-term management or use of depleted uranium hexafluoride. This Request was made to help ensure that, by seeking as many recommendations as possible, Department management considers reasonable options in the long-range management strategy. The Depleted Uranium Hexafluoride Management Program consists of three major program elements: Engineering Analysis, Cost Analysis, and an Environmental Impact Statement. This Technology Assessment Report is the first part of the Engineering Analysis Project, and assesses recommendations from interested persons, industry, and Government agencies for potential uses for the depleted uranium hexafluoride stored at the gaseous diffusion plants in Paducah, Kentucky, and Portsmouth, Ohio, and at the Oak Ridge Reservation in Tennessee. Technologies that could facilitate the long-term management of this material are also assessed. The purpose of the Technology Assessment Report is to present the results of the evaluation of these recommendations. Department management will decide which recommendations will receive further study and evaluation.

With the publication of a Request for Recommendations and Advance Notice of Intent in the November 10, 1994 Federal Register, the Department of Energy initiated a program to assess alternative strategies for the long-term management or use of depleted uranium hexafluoride. This Request was made to help ensure that, by seeking as many recommendations as possible, Department management considers reasonable options in the long-range management strategy. The Depleted Uranium Hexafluoride Management Program consists of three major program elements: Engineering Analysis, Cost Analysis, and an Environmental Impact Statement. This Technology Assessment Report is the first part of the Engineering Analysis Project, and assesses recommendations from interested persons, industry, and Government agencies for potential uses for the depleted uranium hexafluoride stored at the gaseous diffusion plants in Paducah, Kentucky, and Portsmouth, Ohio, and at the Oak Ridge Reservation in Tennessee. Technologies that could facilitate the long-term management of this material are also assessed. The purpose of the Technology Assessment Report is to present the results of the evaluation of these recommendations. Department management will decide which recommendations will receive further study and evaluation. These Appendices contain the Federal Register Notice, comments on evaluation factors, independent technical reviewers resumes, independent technical reviewers manual, and technology information packages.

For many commercial and military applications, the successive Vacuum Induction Melting of uranium metal in graphite crucibles results in a product which is out of specification in carbon. The current recovery method involves dissolution of the metal in acid and chemical purification. This is both expensive and generates mixed waste. A study was undertaken at Lawrence Livermore National Laboratory to investigate the feasibility of reducing the carbon content of uranium metal using electron beam techniques. Results will be presented on the rate and extent of carbon removal as a function of various operating parameters.

A 250 kW electron beam cold hearth refining (EBCHR) melt furnace at Lawrence Livermore National Laboratory (LLNL) has been in operation for over a year producing 5.5 in.-diameter ingots of various uranium alloys. Production of in-specification uranium-6%-niobium (U-6Nb) alloy ingots has been demonstrated using virgin feedstock. A vibratory scrap feeder has been installed on the system and the ability to recycle chopped U-6Nb scrap has been established. A preliminary comparison of vacuum arc remelted (VAR) and electron beam (EB) melted product is presented.

A series of calculations based on critical experiments have been performed using the KENO V.a Monte Carlo Criticality Program for the purpose of validating KENO V.a for use in evaluating Y-12 Plant criticality problems. The experiments were reflected and unreflected systems of single units and arrays containing highly enriched uranium metal or uranium compounds. Various geometrical shapes were used in the experiments. The SCALE control module CSAS25 with the 27-group ENDF/B-4 cross-section library was used to perform the calculations. Some of the experiments were also calculated using the 16-group Hansen-Roach Library. Results are presented in a series of tables and discussed. Results show that the criteria established for the safe application of the KENO IV program may also be used for KENO V.a results.

This Risk Assessment evaluated potential impacts to public health or the environment caused by ground water contamination at the former uranium mill processing site. In the first phase of the U.S. Department of Energy`s Uranium Mill Tailings Remedial Action (UMTRA) Project, the tailing and other contaminated material at this site were placed in a disposal cell near the Gas Hills Plant in 1990. The second phase of the UMTRA Project is to evaluate ground water contamination. This risk assessment is the first site-specific document to evaluate potential health and environmental risks for the Riverton site under the Ground Water Project; it will help determine whether remedial actions are needed for contaminated ground water at the site.

The Idaho National Engineering Laboratory (INEL) is developing new methods to produce high-density aggregate (artificial rock) primarily consisting of depleted uranium oxide. The objective is to develop a low-cost method whereby uranium oxide powder (UO[sub 2], U[sub 3]O[sub ]8, or UO[sub 3]) can be processed to produce high-density aggregate pieces (DU-AGG) having physical properties suitable for disposal in low-level radioactive disposal facilities or for use as a component of high-density concrete used as shielding for radioactive materials. A commercial company, G-M Systems, conducted a design study for a manufacturing pilot plant to process DU-AGG. The results of that study are included and summarized in this report. Also explained are design considerations, equipment capacities, the equipment list, system operation, layout of equipment in the plant, cost estimates, and the proposed plan and schedule.

A radiological survey was conducted at a former uranium recovery plant near Tampa, Florida, operated as a part of a phosphoric acid plant. The uranium recovery operations were conducted from 1951 through 1960, the primary goal being the extraction of uranium from phosphoric acid. Pilot operations were first carried out at a small plant, and full-scale extraction was later carried out at a larger adjacent process plant. The survey included measurement of the followng: beta-gamma dose rates at 1 cm from surfaces and external gamma radiation levels at the surfaces and 1 m above the floor inside the pilot operations building and process building and outdoors in areas around these buildings; fixed and transferable alpha and beta-gamma contamination levels on the floor, walls, ceilings, and roof of the process building and on the floor, walls, and ceiling of the pilot plant offices; concentrations of /sup 226/Ra and /sup 238/U in soil samples taken at grid points around the buildings and in residue samples taken inside the process building; concentrations of /sup 226/Ra and /sup 238/U in water and sediment samples taken outdoors on the site and the concentration of these same nuclides in background samples collected off the site. It was found that beta-gamma and/or alpha contamination levels on surfaces exceed current guidelines for the release of property for unrestricted use at some points inside the process building and in the outdoor area near the process building and pilot operations building. Some samples of soil and residue taken from the floor and equipment on the second level of the process building contained natural uranium in excess of 0.05% by weight and contained natural radium in excess of 900 pCi/g.

This paper reports on macroporous chelating resins (RSP, RSPO, RCSP, and RCSPO) containing dihydroxphosphino and/or -phosphono groups were examined for the concentration and separation of uranium from acidic eluates of macroporous chelating resin containing amidoxime groups. RSP and RSPO had a high adsorption capacity for uranium even in 0.25-0.50 mol {center dot} dm{sup {minus}3} H{sub 2}SO{sub 4}. Uranium adsorbed on the resins was eluted easily as a uranyl carbonate complex by use of 0.25 mol {center dot} dm{sup {minus}3} Na{sub 2}CO{sub 3}. In this effluent, other metal ions were hardly present. The use of RSP and RSPO was very effective in concentrating uranium from seawater and separating it from most other elements.

The mission of this Conversion and Blending Facility (CBF) will be to blend surplus HEU metal and alloy with depleted uranium metal to produce an LEU product. The primary emphasis of this blending operation will be to destroy the weapons capability of large, surplus stockpiles of HEU. The blended LEU product can only be made weapons capable again by the uranium enrichment process. The blended LEU will be produced as a waste suitable for storage or disposal.

The AVLIS Production Plant is designated as a Major System Acquisition (in accordance with DOE Order 4240.IC) to deploy Atomic Vapor Laser Isotope Separation (AVLIS) technology at the Oak Ridge, Tennessee site, in support of the US Uranium Enrichment Program. The AVLIS Production Plant Project will deploy AVLIS technology by performing the design, construction, and startup of a production plant that will meet capacity production requirements of the Uranium Enrichment Program. The AVLIS Production Plant Project Management Plan has been developed to outline plans, baselines, and control systems to be employed in managing the AVLIS Production Plant Project and to define the roles and responsibilities of project participants. Participants will develop and maintain detailed procedures for implementing the management and control systems in agreement with this plan. This baseline document defines the system that measures work performed and costs incurred. This plan was developed by the AVLIS Production Plant Project staff of Martin Marietta Energy Systems, Inc. and Lawrence Livermore National Laboratory in accordance with applicable DOE directives, orders and notices. 38 figures, 19 tables.

The properties of various adsorbents for uranium extraction from seawater are measured under standardized experimental conditions. It turns out that fractionated humic acids have exceptionally fast loading kinetics. This property leads to a substantial reduction of capital investments in conventional adsorbent bed techniques as well as in a procedure designed to avoid large adsorbent bed constructions by using carrier bodies in the open sea.

Uranium contamination of the subsurface remains a persistent problem plaguing remedial design at sites across the U.S. that were involved with production, handling, storage, milling, and reprocessing of uranium for both civilian and defense related purposes. Remediation efforts to date have relied upon excavation, pump-and-treat, or passive remediation barriers (PRB?s) to remove or attenuate uranium mobility. Documented cases convincingly demonstrate that excavation and pump-and-treat methods are ineffective for a number of highly contaminated sites. There is growing concern that use of conventional PRB?s, such as zero-valent iron, may be a temporary solution to a problem that will persist for thousands of years. Alternatives to the standard treatment methods are therefore warranted. The core objective of our research is to demonstrate that a phosphorus amendment strategy will result in a reduction of dissolved uranium to below the proposed drinking water standard. Our hypothesis is that long-chain sodium polyphosphate compounds forestall precipitation of sparingly soluble uranyl phosphate compounds, which is paramount to preventing fouling of wells at the point of injection.

-239 for use in subsequent reactors. A fast neutron reactor is capable of producing more plutonium fuel than the uranium fuel it burns, leading to a breeder reactor. In addition, if the reactor is a fast with half lives of 30 years or less. The fast neutron reactor of preference was to be cooled with liquid

Multiphoton ionization (MPI) time-of-flight mass spectroscopy and photoelectron spectroscopy studies of UF{sub 6} have been conducted using focused light from the Nd:YAG laser fundamental ({lambda}=1064 nm) and its harmonics ({lambda}=532, 355, or 266 nm), as well as other wavelengths provided by a tunable dye laser. The MPI mass spectra are dominated by the singly and multiply charged uranium ions rather than by the UF{sub x}{sup +} fragment ions even at the lowest laser power densities at which signal could be detected. The laser power dependence of U{sup n+} ions signals indicates that saturation can occur for many of the steps required for their ionization. In general, the doubly-charged uranium ion (U{sup 2+}) intensity is much greater than that of the singly-charged uranium ion (U{sup +}). For the case of the tunable dye laser experiments, the U{sup n+} (n = 1- 4) wavelength dependence is relatively unstructured and does not show observable resonance enhancement at known atomic uranium excitation wavelengths. The dominance of the U{sup 2+} ion and the absence or very small intensities of UF{sub x}{sup +} fragments, along with the unsaturated wavelength dependence, indicate that mechanisms may exist other than ionization of bare U atoms after the stepwise photodissociation of F atoms from the parent molecule.

We study Uranium isotopes and surrounding elements at very large neutron number excess. Relativistic mean field and Skyrme-type approaches with different parametrizations are used in the study. Most models show clear indications for isotopes that are stable with respect to neutron emission far beyond N=184 up to the range of around N=258.

Uranium and its associated trace elements and radionuclides are ubiquitous in the South Texas Tertiary environment. Surface mining of this resource from the 1960s through the early 1980s at over sixty locations has left an extensive anthropological footprint (Fig. 1) in the lower Nueces and San Antonio river basins. Reclamation of mining initiated after 1975 has been under the regulatory authority of the Railroad Commission of Texas (RCT). However, mines that were active before the Texas Surface Mining Act of 1975 was enacted, and never reclaimed, are now considered abandoned. The Abandoned Mine Land Section of the RCT is currently reclaiming these pre-regulation uranium mines with funding from the federal government. The RCT monitors the overall effectiveness of this process through post-reclamation radiation and vegetative cover surveys, water quality testing, slope stability and erosion control monitoring. Presently a number of graduate and postgraduate students are completing research on the watershed and reservoir distribution of trace elements and radionuclides downstream of the South Texas Uranium District. The question remains as to whether the elevated levels of uranium, its associated trace elements and radiation levels in the South Texas environment are due to mining

A mineralogical and geochemical study of soils developed from the unmined Coles Hill uranium deposit (Virginia) was undertaken to determine how phosphorous influences the speciation of uranium in an oxidizing soil/saprolite system typical of the eastern United States. This paper presents mineralogical and geochemical results that identify and quantify the processes by which uranium has been sequestered in these soils. It was found that uranium is not leached from the saturated soil zone (saprolites) overlying the deposit due to the formation of a sparingly soluble uranyl phosphate mineral of the meta-autunite group. The concentration of uranium in the saprolites is approximately 1000 mg uranium per kg of saprolite. It was also found that a significant amount of uranium was retained in the unsaturated soil zone overlying uranium-rich saprolites. The uranium concentration in the unsaturated soils is approximately 200 mg uranium per kg of soil (20 times higher than uranium concentrations in similar soils adjacent to the deposit). Mineralogical evidence indicates that uranium in this zone is sequestered by a barium-strontium-calcium aluminum phosphate mineral of the crandallite group (gorceixite). This mineral is intimately inter-grown with iron and manganese oxides that also contain uranium. The amount of uranium associated with both the aluminum phosphates (as much as 1.4 weight percent) has been measured by electron microprobe micro-analyses and the geochemical conditions under which these minerals formed has been studied using thermodynamic reaction path modeling. The geochemical data and modeling results suggest the meta-autunite group minerals present in the saprolites overlying the deposit are unstable in the unsaturated zone soils overlying the deposit due to a decrease in soil pH (down to a pH of 4.5) at depths less than 5 meters below the surface. Mineralogical observations suggest that, once exposed to the unsaturated environment, the meta-autunite group minerals react to form U(VI)- bearing aluminum phosphates. (author)

The main objectives of this work are to determine the hydrogeochemical evolution of an uranium ore and identify the main water/rock interaction processes that control the dissolved uranium content. The Mina Fe uranium-ore deposit is the most important and biggest mine worked in Spain. Sageras area is located at the north part of the Mina Fe, over the same ore deposit. The uranium deposit was not mined in Sageras and was only perturbed by the exploration activities performed 20 years ago. The studied area is located 10 Km northeast of Ciudad Rodrigo (Salamanca) at an altitude over 650 m.a.s.l. The uranium mineralization is related to faults affecting the metasediments of the Upper Proterozoic to Lower Cambrian schist-graywacke complex (CEG), located in the Centro-Iberian Zone of the Hesperian Massif . The primary uranium minerals are uraninite and coffinite but numerous secondary uranium minerals have been formed as a result of the weathering processes: yellow gummite, autunite, meta-autunite, torbernite, saleeite, uranotile, ianthinite and uranopilite. The water flow at regional scale is controlled by the topography. Recharge takes place mainly in the surrounding mountains (Sierra Pena de Francia) and discharge at fluvial courses, mainly Agueda and Yeltes rivers, boundaries S-NW and NE of the area, respectively. Deep flows (lower than 100 m depth) should be upwards due to the river vicinity, with flow directions towards the W, NW or N. In Sageras-Mina Fe there are more than 100 boreholes drilled to investigate the mineral resources of the deposit. 35 boreholes were selected in order to analyze the chemical composition of groundwaters based on their depth and situation around the uranium ore. Groundwater samples come from 50 to 150 m depth. The waters are classified as calcium-bicarbonate type waters, with a redox potential that indicates they are slightly reduced (values vary between 50 to -350 mV). The TOC varies between <0.1 and 4.0 mgC/L and the dissolved uranium has a maximum value of 7.7 mg/L. According the analytical data of dissolved uranium, the mineral closest to equilibrium seems to be UO{sub 2}(am). The tritium contents in the groundwaters vary between 1.5 and 7.3 T.U. Considering that the mean value of tritium in rainwater from the studied area has a value of 4 T.U., it can be concluded that the residence times of the groundwaters are relatively short, not longer than 50 years in the oldest case. (authors)

This document is an environmental assessment of the Spook, Wyoming, Uranium Mill Tailings Remedial Action (UMTRA) Project site. It analyzes the impacts of the U.S. Department of Energy (DOE) proposed action for ground water compliance. The proposed action is to comply with the U.S. Environmental Protection Agency (EPA) standards for the UMTRA Project sites (40 CFR Part 192) by meeting supplemental standards based on the limited use ground water at the Spook site. This proposed action would not require site activities, including ground water monitoring, characterization, or institutional controls. Ground water in the uppermost aquifer was contaminated by uranium processing activities at the Spook site, which is in Converse County, approximately 48 miles (mi) (77 kilometers [km]) northeast of Casper, Wyoming. Constituents from the site infiltrated and migrated into the uppermost aquifer, forming a plume that extends approximately 2500 feet (ft) (800 meters [m]) downgradient from the site. The principal site-related hazardous constituents in this plume are uranium, selenium, and nitrate. Background ground water in the uppermost aquifer at the site is considered limited use. It is neither a current nor a potential source of drinking water because of widespread, ambient contamination that cannot be cleaned up using treatment methods reasonably employed in public water supply systems (40 CFR {section} 192.11 (e)). Background ground water quality also is poor due to first, naturally occurring conditions (natural uranium mineralization associated with an alteration front), and second, the effects of widespread human activity not related to uranium milling operations (uranium exploration and mining activities). There are no known exposure pathways to humans, animals, or plants from the contaminated ground water in the uppermost aquifer because it does not discharge to lower aquifers, to the surface, or to surface water.

Abandoned uranium mill tailings sites in the Four Corners region are a lasting legacy of the Cold War. The U.S. Department of Energy (DOE) is designing landfill repositories that will isolate hazardous constituents of tailings from biological intrusion, erosion, and the underlying aquifer for up to 1,000 years. With evidence of relatively rapid past climate change, and model predictions of global climatic variation exceeding the historical record, DOE recognizes a need to incorporate possible ranges of future climatic and ecological change in the repository design process. In the Four Corners region, the center of uranium mining and milling activities in the United States, proxy paleoclimatic records may be useful not only as a window on the past, but also as analogs of possible local responses to future global change. We reconstructed past climate change using available proxy data from tree rings, packrat middens, lake sediment pollen, and archaeological records. Interpretation of proxy paleoclimatic records was based on present-day relationships between plant distribution, precipitation, and temperature along a generalized elevational gradient for the region. For the Monticello, Utah, uranium mill tailings site, this first approximation yielded mean annual temperature and precipitation ranges of 2 to 10{degrees} C, and 38 to 80 cm, respectively, corresponding to late glacial and Altithermal periods. These data are considered to be reasonable ranges of future climatic conditions that can be input to evaluations of groundwater recharge, radon-gas escape, erosion, frost penetration, and biointrusion in engineered earthen barriers designed to isolate tailings.

The path forward for managing of Hanford K Basin sludge calls for it to be packaged, shipped, and stored at T Plant until final processing at a future date. An important consideration for the design and cost of retrieval, transportation, and storage systems is the potential for heat and gas generation through oxidation reactions between uranium metal and water. This report, the third in a series (Series III), describes work performed at the Pacific Northwest National Laboratory (PNNL) to assess corrosion and gas generation from irradiated metallic uranium particles (fuel particles) with and without K Basin sludge addition. The testing described in this report consisted of 12 tests. In 10 of the tests, 4.3 to 26.4 g of fuel particles of selected size distribution were placed into 60- or 800-ml reaction vessels with 0 to 100 g settled sludge. In another test, a single 3.72-g fuel fragment (i.e., 7150-mm particle) was placed in a 60 ml reaction vessel with no added sludge. The twelfth test contained only sludge. The fuel particles were prepared by crushing archived coupons (samples) from an irradiated metallic uranium fuel element. After loading the sludge materials (whether fuel particles, mixtures of fuel particles and sludge, or sludge-only) into reaction vessels, the solids were covered with an excess of K Basin water, the vessels closed and connected to a gas measurement manifold, and the vessels back-flushed with inert neon cover gas. The vessels were then heated to a constant temperature. The gas pressures and temperatures were monitored continuously from the times the vessels were purged. Gas samples were collected at various times during the tests, and the samples analyzed by mass spectrometry. Data on the reaction rates of uranium metal fuel particles with water as a function of temperature and particle size were generated. The data were compared with published studies on metallic uranium corrosion kinetics. The effects of an intimate overlying sludge layer (''blanket'') on the uranium metal corrosion rates were also evaluated.

Tannin compounds having multiple adjacent hydroxy groups have an extremely high affinity for uranium. To prevent the leaching of tannins into water and to improve the adsorbing characteristics of these compounds, the authors tried to immobilize tannins. The immobilized tannin has the most favorable features for uranium recovery; high selective adsorption ability to uranium, rapid adsorption rate, and applicability in both column and batch systems. The immobilized tannin can recover uranium from natural seawater with high efficiency. About 2530 ..mu..g uranium is adsorbed per gram of this adsorbent within 22 h. Depending on the concentration in seawater, an enrichment of up to 766,000-fold within the adsorbent is possible. Almost all uranium adsorbed is easily desorbed with a very dilute acid. Thus, the immobilized tannin can be used repeatedly in the adsorption-desorption process.

An in-line assay monitor for determining the content of uranium-235 in a uranium hexafluoride gas isotopic separation system is provided which removes the necessity of complete access to the operating parameters of the system for determining the uranium-235 content. The method and monitor for carrying out the method involve cooling of a radiation pervious chamber connected in fluid communication with the selected point in the system to withdraw a specimen and solidify the specimen in the chamber. The specimen is irradiated by means of an ionizing radiation source of energy different from that of the 185 keV gamma emissions from uranium-235. The uranium-235 content of the specimen is determined from comparison of the accumulated 185 keV energy counts and reference energy counts. The latter is used to measure the total uranium isotopic content of the specimen.

The Department of Energy (DOE) is in the process of converting about 700,000 metric tons (MT) of depleted uranium hexafluoride (DUF6) containing 475,000 MT of depleted uranium (DU) to a stable form more suitable for long-term storage or disposal. Potential conversion forms include the tetrafluoride (DUF4), oxide (DUO2 or DU3O8), or metal. If worthwhile beneficial uses cannot be found for the DU product form, it will be sent to an appropriate site for disposal. The DU products are considered to be low-level waste (LLW) under both DOE orders and Nuclear Regulatory Commission (NRC) regulations. The objective of this study was to assess the acceptability of the potential DU conversion products at potential LLW disposal sites to provide a basis for DOE decisions on the preferred DU product form and a path forward that will ensure reliable and efficient disposal.

The objective of this paper is to investigate the transport of uranium oxide particles that may be present in carbon dioxide (CO2) gas coolant, into the graphite blocks of gas-cooled, graphite moderated reactors. The transport of uranium oxide in the coolant system, and subsequent deposition of this material in the graphite, of such reactors is of interest because it has the potential to influence the application of the Graphite Isotope Ratio Method (GIRM). The GIRM is a technology that has been developed to validate the declared operation of graphite moderated reactors. GIRM exploits isotopic ratio changes that occur in the impurity elements present in the graphite to infer cumulative exposure and hence the reactor’s lifetime cumulative plutonium production. Reference Gesh, et. al., for a more complete discussion on the GIRM technology.

This document was prepared to serve as a guide for export control officials in their interpretation, understanding, and implementation of export laws that relate to the Zangger International Trigger List for gaseous diffusion uranium enrichment process components, equipment, and materials. Particular emphasis is focused on items that are especially designed or prepared since export controls are required for these by States that are party to the International Nuclear Nonproliferation Treaty.

The energy return on investment (EROI) of an energy resource is the ratio of the energy it ultimately produces to the energy used to recover it. EROI is a key viability measure for a new recovery technology, particularly in its early stages of development when financial cost assessment would be premature or highly uncertain. This paper estimates the EROI of uranium recovery from seawater via a braid adsorbent technology. In this paper, the energy cost of obtaining uranium from seawater is assessed by breaking the production chain into three processes: adsorbent production, adsorbent deployment and mooring, and uranium elution and purification. Both direct and embodied energy inputs are considered. Direct energy is the energy used by the processes themselves, while embodied energy is used to fabricate their material, equipment or chemical inputs. If the uranium is used in a once-through fuel cycle, the braid adsorbent technology EROI ranges from 12 to 27, depending on still-uncertain performance and system design parameters. It is highly sensitive to the adsorbent capacity in grams of U captured per kg of adsorbent as well as to potential economies in chemical use. This compares to an EROI of ca. 300 for contemporary terrestrial mining. It is important to note that these figures only consider the mineral extraction step in the fuel cycle. At a reference performance level of 2.76 g U recovered per kg adsorbent immersed, the largest energy consumers are the chemicals used in adsorbent production (63%), anchor chain mooring system fabrication and operations (17%), and unit processes in the adsorbent production step (12%). (authors)

Results are presented of a 10 weeks field experiment producing uranium from seawater by the so-called adsorber-loop-concept. For the adsorption process polyamidoxin (PAO) granulate has been used with grain sizes between 0.3 - 1.2 mm diameter. The performance of the adsorber and the efficiency of the adsorption process - especially with regard to high volume flows of seawater - are presented.

Uranium contamination of the subsurface has remained a persistent problem plaguing remedial design at sites across the U.S. that were involved with production, handling, storage, milling, and reprocessing of fissile uranium for both civilian and defense related purposes. Remediation efforts to date have relied upon excavation, pump-and-treat, or passive remediation barriers (PRB?s) to remove or attenuate uranium mobility. Documented cases convincingly demonstrate that excavation and pump-and-treat methods are ineffective for a number of highly contaminated sites. There is growing concern that use of conventional PRB?s, such as zero-valent iron, are a temporary solution to a problem that will persist for thousands of years. Alternatives to the standard treatment methods are therefore warranted. The core objective of our research is to demonstrate that a phosphorous amendment strategy will result in a reduction of dissolved uranium to below the proposed drinking water standard. Our hypothesis is that long-chain polyphosphate compounds forestall precipitation of sparingly soluble uranyl phosphate compounds, which is key to preventing fouling of wells at the point of injection. Our other fundamental objective is to synthesize and correctly characterize the uranyl phosphate phases that form in the geochemical conditions under consideration. This report summarizes work conducted at the University of Notre Dame through November of 2003 under DOE grant DE-FG07-02ER63489, which has been funded since September, 2002. The objectives at Notre Dame are development of synthesis techniques for uranyl phosphate phases, together with detailed structural and chemical characterization of the myriad of uranyl phosphate phases that may form under geochemical conditions under consideration.

who have provided technical support for this project throughout its duration. I also express my sincere appreciation and thanks to the U. S. Department of Energy and Oak Ridge Associated Universities whose Nuclear Engineering and Health Physics... space nuclear reactors. Uranium mononitride is currently the reference fuel for the SP-100 space reactor system and will likely be considered for application in future multimegawatt space power systems as well. Although fuel modeling efforts have...

In this paper we develop a simple model of the penetration process of a long rod through an uniform target. Applying the momentum and energy conservation laws, we derive an analytical relation which shows how the penetration depth depends upon the density of the rod, given a fixed kinetic energy. This work was sparked off by the necessity of understanding the effectiveness of high density penetrators (e.g. depleted uranium penetrators) as anti-tank weapons.

This document provides the documented safety analysis (DSA) and Central Plateau Remediation Project (CP) requirements that apply to surveillance and maintenance (S&M) activities at the Plutonium-Uranium Extraction (PUREX) facility. This DSA was developed in accordance with DOE-STD-1120-98, ''Integration of Environment, Safety, and Health into Facility Disposition Activities''. Upon approval and implementation of this document, the current safety basis documents will be retired.

as uranium concentrations dzop to an average of 2. dppm, indicative of relatively low uranium bearing pelagic particle deposition. Furthermore, the 13 C values become heavier in this region, lacking a large terrest. rial component. This evidence suggests... 39 the basin walls. It may be possible that these particles at the brine/seawater interface incorporate any uranium reduced in this zone and carry it to the brine perimeter where it is deposited. Unfortunately, attempts to accurately core...

Two stratigraphic units, the Late Jurassic Salt Wash Member of the Morrison Formation and the Triassic Chinle Formation, were determined to be favorable for the occurrence of uranium deposits that meet the minimum size and grade requirements of the US Department of Energy in the Salina 1 x 2/sup 0/ Quadrangle, Utah. Three areas judged favorable for the Salt Wash Member are the Tidwell and Notom districts, and the Henry Mountains mineral belt. The criteria used to establish favorability were the presence of: (1) fluvial sandstone beds deposited by low-energy streams; (2) actively moving major and minor structures such as the Paradox basin and the many folds within it; (3) paleostream transport directions approximately perpendicular to the trend of many of the paleofolds; (4) presence of favorable gray lacustrine mudstone beds; and (5) known uranium occurrences associated with the favorable gray mudstones. Four favorable areas have been outlined for the Chinle Formation. These are the San Rafael Swell, Inter River, and the Orange Cliffs subareas and the Capitol Reef area. The criteria used to establish these areas are: the sandstone-to-mudstone ratios and the geographic distribution of the Petrified Forest Member of the Chinle Formation which is considered as the probable source for the uranium.

3-1 Chapter 3. Volume and Characteristics of Uranium Mine Wastes Uranium has been found and mined conventional mining, solution extraction, and milling of uranium, a principal focus of this report is TENORM, or which may need future reclamation. When uranium mining first started, most of the ores were recovered

Recent work at Oak Ridge National Laboratory (ORNL) has focused on some source term modeling of uranyl nitrate (UN) as part of a comprehensive validation effort employing gamma-ray detector instrumentation for the detection of diversion from declared conversion activities. Conversion, the process by which natural uranium ore (yellowcake) is purified and converted through a series of chemical processes into uranium hexafluoride gas (UF6), has historically been excluded from the nuclear safeguards requirements of the 235U-based nuclear fuel cycle. The undeclared diversion of this product material could potentially provide feedstock for a clandestine weapons program for state or non-state entities. Given the changing global political environment and the increased availability of dual-use nuclear technology, the International Atomic Energy Agency has evolved its policies to emphasize safeguarding this potential feedstock material in response to dynamic and evolving potential diversion pathways. To meet the demand for instrumentation testing at conversion facilities, ORNL developed the Uranyl Nitrate Calibration Loop Equipment (UNCLE) facility to simulate the full-scale operating conditions of a purified uranium-bearing aqueous stream exiting the solvent extraction process in a natural uranium conversion plant. This work investigates gamma-ray signatures of UN circulating in the UNCLE facility and evaluates detector instrumentation sensitivity to UN for safeguards applications. These detector validation activities include assessing detector responses to the UN gamma-ray signatures for spectrometers based on sodium iodide, lanthanum bromide, and germanium detectors. The results of measurements under static and dynamic operating conditions at concentrations ranging from 10-90g U/L of naturally enriched UN will be presented. A range of gamma-ray lines was examined and self-attenuation factors were calculated, in addition to attenuation for transmission measurement of density, concentration and enrichment. A detailed uncertainty analysis will be presented providing insights into instrumentation limitations to spoofing.

For nuclear energy to remain sustainable in the United States, economically viable sources of uranium beyond terrestrial ores must be developed. The goal of this program is to develop advanced adsorbents that can extract uranium from seawater at twice the capacity of the best adsorbent developed by researchers at the Japan Atomic Energy Agency (JAEA), 1.5 mg U/g adsorbent. A multidisciplinary team from Oak Ridge National Laboratory, Lawrence Berkeley National Laboratory, Pacific Northwest National Laboratory, and the University of Texas at Austin was assembled to address this challenging problem. Polymeric adsorbents, based on the radiation grafting of acrylonitrile and methacrylic acid onto high surface-area polyethylene fibers followed by conversion of the nitriles to amidoximes, have been developed. These poly(acrylamidoxime-co-methacrylic acid) fibers showed uranium adsorption capacities for the extraction of uranium from seawater that exceed 3 mg U/g adsorbent in testing at the Pacific Northwest National Laboratory Marine Sciences Laboratory. The essence of this novel technology lies in the unique high surface-area trunk material that considerably increases the grafting yield of functional groups without compromising its mechanical properties. This technology received an R&D100 Award in 2012. In addition, high surface area nanomaterial adsorbents are under development with the goal of increasing uranium adsorption capacity by taking advantage of the high surface areas and tunable porosity of carbon-based nanomaterials. Simultaneously, de novo structure-based computational design methods are being used to design more selective and stable ligands and the most promising candidates are being synthesized, tested and evaluated for incorporation onto a support matrix. Fundamental thermodynamic and kinetic studies are being carried out to improve the adsorption efficiency, the selectivity of uranium over other metals, and the stability of the adsorbents. Understanding the rate-limiting step of uranium uptake from seawater is also essential in designing an effective uranium recovery system. Finally, economic analyses have been used to guide these studies and highlight what parameters, such as capacity, recyclability, and stability, have the largest impact on the cost of extraction of uranium from seawater. Initially, the cost estimates by the JAEA for extraction of uranium from seawater with braided polymeric fibers functionalized with amidoxime ligands were evaluated and updated. The economic analyses were subsequently updated to reflect the results of this project while providing insight for cost reductions in the adsorbent development through “cradle-to-grave” case studies for the extraction process. This report highlights the progress made over the last three years on the design, synthesis, and testing of new materials to extract uranium for seawater. This report is organized into sections that highlight the major research activities in this project: (1) Chelate Design and Modeling, (2) Thermodynamics, Kinetics and Structure, (3) Advanced Polymeric Adsorbents by Radiation Induced Grafting, (4) Advanced Nanomaterial Adsorbents, (5) Adsorbent Screening and Modeling, (6) Marine Testing, and (7) Cost and Energy Assessment. At the end of each section, future research directions are briefly discussed to highlight the challenges that still remain to reduce the cost of extractions of uranium for seawater. Finally, contributions from the Nuclear Energy University Programs (NEUP), which complement this research program, are included at the end of this report.

Uranium contaminated soils from the Fernald Operation Site, Ohio, have been examined by a combination of optical microscopy, scanning electron microscopy with backscattered electron detection (SEM/BSE), and analytical electron microscopy (AEM). A method is described for preparing of transmission electron microscopy (TEM) thin sections by ultramicrotomy. By using these thin sections, SEM and TEM images can be compared directly. Uranium was found in iron oxides, silicates (soddyite), phosphates (autunites), and fluorite. Little uranium was associated with clays. The distribution of uranium phases was found to be inhomogeneous at the microscopic level.

An apparatus and method are disclosed for continuous production of liquid uranium alloys through the electrolytic reduction of uranium chlorides. The apparatus includes an electrochemical cell formed from an anode shaped to form an electrolyte reservoir, a cathode comprising a metal, such as iron, capable of forming a eutectic uranium alloy having a melting point less than the melting point of pure uranium, and molten electrolyte in the reservoir comprising a chlorine or fluorine containing salt and uranium chloride. The method of the invention produces an eutectic uranium alloy by creating an electrolyte reservoir defined by a container comprising an anode, placing an electrolyte in the reservoir, the electrolyte comprising a chlorine or fluorine containing salt and uranium chloride in molten form, positioning a cathode in the reservoir where the cathode comprises a metal capable of forming an uranium alloy having a melting point less than the melting point of pure uranium, and applying a current between the cathode and the anode. 2 figures.

Gamma-spectrometric methods suitable for the characterization of highly enriched uranium samples encountered in illicit trafficking of nuclear materials are presented. In particular, procedures for determining the 234U, 235U, 238U, 232U and 236U contents and the age of highly enriched uranium are described. Consequently, the total uranium content and isotopic composition can be calculated. For determining the 238U and 232U contents a low background chamber was used. In addition, age dating of uranium was also performed using low-background spectrometry.

Gamma-spectrometric methods suitable for the characterization of highly enriched uranium samples encountered in illicit trafficking of nuclear materials are presented. In particular, procedures for determining the 234U, 235U, 238U, 232U and 236U contents and the age of highly enriched uranium are described. Consequently, the total uranium content and isotopic composition can be calculated. For determining the 238U and 232U contents a low background chamber was used. In addition, age dating of uranium was also performed using low-background spectrometry.

Methods of fabricating a uranium-bearing foil are described. The foil may be substantially pure uranium, or may be a uranium alloy such as a uranium-molybdenum alloy. The method typically includes a series of hot rolling operations on a cast plate material to form a thin sheet. These hot rolling operations are typically performed using a process where each pass reduces the thickness of the plate by a substantially constant percentage. The sheet is typically then annealed and then cooled. The process typically concludes with a series of cold rolling passes where each pass reduces the thickness of the plate by a substantially constant thickness amount to form the foil.

The authors present results of an investigation of uranium/calcium ratios in cleaned foraminiferal calcite as a recorder of seawater uranium concentrations. For accurate reconstruction of past seawater uranium content, shell calcite must incorporate uranium in proportion to seawater concentration and must preserve its original uranium composition over time. Laboratory culture experiments with live benthic (Amphistegina lobifera) and live planktonic (Globigerinell calida) foraminifera show that the U/Ca ratio of cleaned calcite tests is proportional to the concentration of uranium in solution. After correcting results for the presence of initial calcite, the apparent distribution coefficient D = (U/Ca[sub calcite])/(U/Ca)[sub solution] = 10.6 [+-] 0.3 (x10[sup [minus]3]) for A. lobifera and D = 7.9 [+-] 0.1 (x10[sup [minus]3]) for G. calida. U/Ca ratios in planktonic foraminifera from core tops collected above 3900 m in the equatorial Atlantic and above 2100 m in the Pacific Ocean show no significant difference among the species analyzed. D estimated form core top samples ranges from 7.6 [+-] 0.4 (x10[sup [minus]3]) for O. universa to 8.4 [+-] 0.5 (x10[sup [minus]3]) for G. ruber. In benthic species C. wuellerstorfi, D = 7.0 [+-] 0.8 (x10[sup [minus]3]). U/Ca and Mg/Ca in G. tumida and G. sacculifer from core tops taken near and below the regional lysocline decrease with water depth. Smaller decreases in U/Ca and Mg/Ca with depth were observed in C. wuellerstorfi. In the planktonic species, the authors believe that U/CA and Mg/Ca are lower in the more dissolution-resistant fraction of calcite, leading to lower U/Ca in more highly dissolved samples.

During 2009 nuclear power plants, with a capacity of 370 GWe, will produce roughly 14% of the worldwide electric energy. About 65000 tons of natural uranium equivalent are required to operate these reactors. For 15 years on average only 2/3 of this fuel is provided by the uranium mines and 1/3 comes from secondary resources. In this paper the situation concerning the secondary resources at the beginning of the year 2009 is presented. The data used are from the IAEA/NEA 2007 Red Book, "Uranium Resources, Production and Demand", and from the World Nuclear Association (WNA). Our analysis shows that these civilian stocks will be essentially exhausted within the next 5 years. This coincides roughly with the year 2013, when the delivery of the 10000 tons of natural uranium equivalent from russian military stocks to the USA will end. As the majority of the remaining civilian stocks, about 30000 tons, are believed to be under the control of the US government and american companies, it seems rather unlikely that the U...

The Waste Isolation Pilot Plant (WIPP) is being considered for the disposal of K Basin sludge as RH-TRU. Because the hydrogen gas concentration in the 55-gallon RH-TRU sealed drums to be transported to WIPP is limited by flammability safety, the number of containers and shipments likely will be driven by the rate of hydrogen generated by the uranium metal-water reaction (U + 2 H{sub 2}O {yields} UO{sub 2} + 2 H{sub 2}) in combination with the hydrogen generated from water and organic radiolysis. Gas generation testing was conducted with uranium metal particles of known surface area, in simulated K West (KW) Basin canister sludge and immobilized in candidate grout solidification matrices. This study evaluated potential for Portland cement and magnesium phosphate grouts to inhibit the reaction of water with uranium metal in the sludge and thereby permit higher sludge loading to the disposed waste form. The best of the grouted waste forms decreased the uranium metal-water reaction by a factor of four.

Historically, the approach to safeguarding nuclear material in the front end of the fuel cycle was implemented only at the stage when UF6 was declared as feedstock for enrichment plants. Recent International Atomic Energy Agency (IAEA) circulars and policy papers have sought to implement safeguards when any purified aqueous uranium solution or uranium oxides suitable for isotopic enrichment or fuel fabrication exist. Oak Ridge National Laboratory has developed the Uranyl Nitrate Calibration Loop Equipment (UNCLE) facility to simulate the full-scale operating conditions for a purified uranium-bearing aqueous stream exiting the solvent extraction process conducted in a natural uranium conversion plant (NUCP) operating at 6000 MTU/year. Monitoring instruments, including the 3He passive neutron detector developed at Los Alamos National Laboratory and the Endress+Hauser Promass 83F Coriolis meter, have been tested at UNCLE and field tested at Springfields. The field trials demonstrated the need to perform full-scale equipment testing under controlled conditions prior to field deployment of operations and safeguards monitoring at additional plants. Currently, UNCLE is testing neutron-based monitoring for detection of noncompliant activities; however, gamma-ray source term monitoring is currently being explored complementary to the neutron detector in order to detect undeclared activities in a more timely manner. The preliminary results of gamma-ray source term modeling and monitoring at UNCLE are being analyzed as part of a comprehensive source term and detector benchmarking effort. Based on neutron source term detection capabilities, alternative gamma-based detection and monitoring methods will be proposed to more effectively monitor NUCP operations in verifying or detecting deviations from declared conversion activities.

As part of the National Nuclear Security Agency (NNSA) mandate to replace the use of highly enriched uranium (HEU) fuel for low enriched uranium (LEU) fuel, research into the development of LEU fuel for research reactors has been active since the late 1970’s. Originally referred to as the Reduced Enrichment for Research and Test Reactor (RERTR) program the new effort named Global Threat Reduction Initiative (GTRI) is nearing the goal of replacing the standard aluminum clad dispersion highly enriched uranium aluminide fuel with a new LEU fuel. The five domestic high performance research reactors undergoing this conversion are High Flux Isotope reactor (HFIR), Advanced Test Reactor (ATR), National Institute of Standards and Technology (NIST) Reactor, Missouri University Research Reactor (MURR) and the Massachusetts Institute of Technology Reactor II (MITR-II). The design of these reactors requires a higher neutron flux than other international research reactors, which to this point has posed unique challenges in the design and development of the new mandated LEU fuel. The new design utilizes a monolithic fuel configuration in order to obtain sufficient 235U within the LEU stoichoimetry to maintain the fission reaction within the domestic test reactors. The change from uranium aluminide dispersion fuel type to uranium molybdenum (UMo) monolithic configuration requires examination of possible corrosion issues associated with the new fuel meat. A focused analysis of the UMo fuel under potential corrosion conditions, within the ATR and under aqueous storage indicates a slow and predictable corrosion rate. Additional corrosion testing is recommended for the highest burn-up fuels to confirm observed corrosion rate trends. This corrosion analysis will focus only on the UMo fuel and will address corrosion of ancillary components such as cladding only in terms of how it affects the fuel. The calculations and corrosion scenarios are weighted with a conservative bias to provide additional confidence with the results. The actual corrosion rates of UMo fuel is very likely to be lower than assumed within this report which can be confirmed with additional testing.

The invention is a process for the recovery of uranium from uranium-bearing hydrocarbon oils containing carboxylic acid as a degradation product. In one aspect, the invention comprises providing an emulsion of water and the oil, heating the same to a temperature effecting conversion of the emulsion to an organic phase and to an acidic aqueous phase containing uranium carboxylate, and recovering the uranium from the aqueous phase. The process is effective, simple and comparatively inexpensive. It avoids the use of toxic reagents and the formation of undesirable intermediates.

This report contains a cost analysis of the long-term storage of depleted uranium in the form of uranium metal. Three options are considered for storage of the depleted uranium. These options are aboveground buildings, partly underground vaults, and mined cavities. Three cases are presented. In the first case, all the depleted uranium metal that would be produced from the conversion of depleted uranium hexafluoride (UF{sub 6}) generated by the US Department of Energy (DOE) prior to July 1993 would be stored at the storage facility (100% Case). In the second case, half the depleted uranium metal would be stored at this storage facility (50% Case). In the third case, one-quarter of the depleted uranium metal would be stored at the storage facility (25% Case). The technical basis for the cost analysis presented in this report is principally found in the companion report, ANL/EAD/TM-100, ''Engineering Analysis Report for the Long-Term Management of Depleted Uranium Hexafluoride: Storage of Depleted Uranium Metal'', prepared by Argonne National Laboratory.

A complex approach based on the consistent modeling of neutron-physics processes and processes of cascade separation of isotopes is applied for analyzing physical problems of the multiple usage of reprocessed uranium in the fuel cycle of light water reactors. A number of scenarios of multiple recycling of reprocessed uranium in light water reactors are considered. In the process, an excess absorption of neutrons by the {sup 236}U isotope is compensated by re-enrichment in the {sup 235}U isotope. Specific consumptions of natural uranium for re-enrichment of the reprocessed uranium depending on the content of the {sup 232}U isotope are obtained.

Quantification of the residual nuclear material remaining in process equipment has long been a challenge to those who work with nuclear material accounting systems. Fortunately, nuclear material has spontaneous radiation emissions that can be measured. If gamma-ray measurements can be made, it is easy to determine what isotope a deposit contains. Unfortunately, it can be quite difficult to relate this measured signal to an estimate of the mass of the nuclear deposit. Typically, the measurement expert must work with incomplete or inadequate information to determine a quantitative result. Simplified analysis models, the distribution of the nuclear material, any intervening attenuation, background(s), and the source-to-detector distance(s) can have significant impacts on the quantitative result. This presentation discusses the application of a generalized-geometry holdup model to the low-enriched uranium fuel pellet fabrication plant in Ust-Kamenogorsk, Kazakhstan. Preliminary results will be presented. Software tools have been developed to assist the facility operators in performing and documenting the measurements. Operator feedback has been used to improve the user interfaces.

As the uranium resource in terrestrial ores is limited, it is difficult to ensure a long-term sustainable nuclear energy technology. The oceans contain approximately 4.5 billion tons of uranium, which is one thousand times the amount of uranium in terrestrial ores. Development of technologies to recover the uranium from seawater would greatly improve the uranium resource availability, sustaining the fuel supply for nuclear energy. Several methods have been previously evaluated including solvent extraction, ion exchange, flotation, biomass collection, and adsorption; however, none have been found to be suitable for reasons such as cost effectiveness, long term stability, and selectivity. Recent research has focused on the amidoxime functional group as a promising candidate for uranium sorption. Polymer beads and fibers have been functionalized with amidoxime functional groups, and uranium adsorption capacities as high as 1.5 g U/kg adsorbent have recently been reported with these types of materials. As uranium concentration in seawater is only ~3 ppb, great improvements to uranium collection systems must be made in order to make uranium extraction from seawater economically feasible. This proposed research intends to develop transformative technologies for economic uranium extraction from seawater. The Lin group will design advanced porous supports by taking advantage of recent breakthroughs in nanoscience and nanotechnology and incorporate high densities of well-designed chelators into such nanoporous supports to allow selective and efficient binding of uranyl ions from seawater. Several classes of nanoporous materials, including mesoporous silica nanoparticles (MSNs), mesoporous carbon nanoparticles (MCNs), meta-organic frameworks (MOFs), and covalent-organic frameworks (COFs), will be synthesized. Selective uranium-binding liagnds such as amidoxime will be incorporated into the nanoporous materials to afford a new generation of sorbent materials that will be evaluated for their uranium extraction efficiency. The initial testing of these materials for uranium binding will be carried out in the Lin group, but more detailed sorption studies will be carried out by Dr. Taylor-Pashow of Savannah River National Laboratory in order to obtain quantitative uranyl sorption selectivity and kinetics data for the proposed materials. The proposed nanostructured sorbent materials are expected to have higher binding capacities, enhanced extraction kinetics, optimal stripping efficiency for uranyl ions, and enhanced mechanical and chemical stabilities. This transformative research will significantly impact uranium extraction from seawater as well as benefit DOE’s efforts on environmental remediation by developing new materials and providing knowledge for enriching and sequestering ultralow concentrations of other metals.

An extensive series of delayed critical experiments were performed at the Oak Ridge Critical Experiments Facility using enriched uranium metal during the 1960s and 1970s in support of criticality safety operations at the Y-12 Plant. These experiments were designed to evaluate the storage, casting, and handling limits of the Y-12 Plant and to provide data for the verification of cross sections and calculation methods utilized in nuclear criticality safety applications. Many of these experiments have already been evaluated and included in the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbook: unreflected (HEU-MET-FAST-051), graphite-reflected (HEU-MET-FAST-071), and polyethylene-reflected (HEU-MET-FAST-076). Three of the experiments consisted of highly-enriched uranium (HEU, ~93.2% 235U) metal parts reflected by beryllium metal discs. The first evaluated experiment was constructed from a stack of 7-in.-diameter, 4-1/8-in.-high stack of HEU discs top-reflected by a 7-in.-diameter, 5-9/16-in.-high stack of beryllium discs. The other two experiments were formed from stacks of concentric HEU metal annular rings surrounding a 7-in.diameter beryllium core. The nominal outer diameters were 13 and 15 in. with a nominal stack height of 5 and 4 in., respectively. These experiments have been evaluated for inclusion in the ICSBEP Handbook.

The reaction of uranium oxides and chlorine to prepare anhydrous uranium tetrachloride (UCl{sub 4}) are important to more economical preparation of uranium metal. The most practical reactions require carbon or carbon monoxide (CO) to give CO or carbon dioxide (CO{sub 2}) as waste gases. The chemistry of U-O-Cl compounds is very complex with valances of 3, 4, 5, and 6 and with stable oxychlorides. Literature was reviewed to collect thermochemical data, phase equilibrium information, and results of experimental studies. Calculations using thermodynamic data can identify the probable reactions, but the results are uncertain. All the U-O-Cl compounds have large free energies of formation and the calculations give uncertain small differences of large numbers. The phase diagram for UCl{sub 4}-UO{sub 2} shows a reaction to form uranium oxychloride (UOCl{sub 2}) that has a good solubility in molten UCl{sub 4}. This appears more favorable to good rates of reaction than reaction of solids and gases. There is limited information on U-O-Cl salt properties. Information on the preparation of titanium, zirconium, silicon, and thorium tetrachlorides (TiCl{sub 4}, ZrCl{sub 4}, SiCl{sub 4}, ThCl{sub 4}) by reaction of oxides with chlorine (Cl{sub 2}) and carbon has application to the preparation of UCl{sub 4}.

Uranium deposits in the South Texas Uranium Region are classical roll-type deposits that formed at the margin of tongues of altered sandstone by the encroachment of oxidizing, uraniferous solutions into reduced aquifers containing pyrite and, in a few cases, carbonaceous plant material. Many of the uranium deposits in South Texas are dissimilar from the roll fronts of the Wyoming basins. The host sands for many of the deposits contain essentially no carbonaceous plant material, only abundant disseminated pyrite. Many of the deposits do not occur at the margin of altered (ferric oxide-bearing) sandstone tongues but rather occur entirely within reduced, pyurite-bearing sandstone. The abundance of pyrite within the sands probably reflects the introduction of H/sub 2/S up along faults from hydrocarbon accumulations at depth. Such introductions before ore formation prepared the sands for roll-front development, whereas post-ore introductions produced re-reduction of portions of the altered tongue, leaving the deposit suspended in reduced sandstone. Evidence from three deposits suggests that ore formation was not accompanied by the introduction of significant amounts of H/sub 2/S.

As part of a global effort to convert reactors that require highly enriched uranium to instead operate with low enriched uranium, monolithic fuel plates consisting of a U-Mo fuel meat with a zirconium foil barrier layer and clad in aluminum...

NUREG/CR-6911 Tests of Uranium (VI) Adsorption Models in a Field Setting U.S. Geological Survey U/CR-6911 Tests of Uranium (VI) Adsorption Models in a Field Setting Manuscript Completed: August 2006 Date observations clearly demonstrated that in aquifers where U(VI) concentrations are controlled by adsorption

Uranium is removed from the oceans by diffusion across the sediment-water interface of organic-rich sediments. This pathway is the largest single sink in the global budget of this element. Dissolved uranium is drawn into suboxic sediments along a concentration gradient established by the precipitation of an insoluble phase which forms when U(VI) is reduced to U(IV). This transformation occurs relatively late in the diagenetic sequence, after the microbially mediated dissolution of manganese and iron oxides, and may be induced by the onset of sulfate reduction. Metallo-organics play an important role in the diagenetic behavior of this element as some uranium is released into solution when labile organics are consumed at the sediment-water interface. In contrast, the diagenesis of authigenic Fe- and Mn-oxides exerts negligible influence on the uranium diagenetic cycle. Variations in the uranium concentration of sediment with time are controlled directly by the uranium content of the source material settling from the water column, and indirectly, by the organic content of this material and sedimentation rate. Since diffusion from seawater influences dramatically the short-term burial rate of uranium, down-core distributions of dissolved and solid uranium can provide an estimate of recent sedimentation rates in rapidly accumulating sediments.

Current light water reactors (LWR) burn fissile uranium, whereas some future reactors, as Sodium fast reactors (SFR) will be capable of recycling their own plutonium and already-extracted depleted uranium. This makes them a feasible solution for the sustainable development of nuclear energy. Nonetheless, a sufficient quantity of plutonium is needed to start up an SFR, with the plutonium already being produced in light water reactors. The availability of natural uranium therefore has a direct impact on the capacity of the reactors (both LWR and SFR) that we can build. It is therefore important to have an accurate estimate of the available uranium resources in order to plan for the world's future nuclear reactor fleet. This paper discusses the correspondence between the resources (uranium and plutonium) and the nuclear power demand. Sodium fast reactors will be built in line with the availability of plutonium, including fast breeders when necessary. Different assumptions on the global uranium resources are taken into consideration. The largely quoted estimate of 22 Mt of uranium recovered for phosphate rocks can be seriously downscaled. Based on our current knowledge of phosphate resources, 4 Mt of recoverable uranium already seems to be an upper bound value. The impact of the downscaled estimate on the deployment of a nuclear fleet is assessed accordingly. (authors)

Niobium is recovered from scrap uranium-niobium alloy by melting the scrap with tin, solidifying the billet thus formed, heating the billet to combine niobium with tin therein, placing the billet in hydrochloric acid to dissolve the uranium and leave an insoluble residue of niobium stannide, then separating the niobium stannide from the acid.

Niobium is recovered from scrap uranium-niobium alloy by melting the scrap with tin, solidifying the billet thus formed, heating the billet to combine niobium with tin therein, placing the billet in hydrochloric acid to dissolve the uranium and leave an insoluble residue of niobium stannide, then separating the niobium stannide from the acid.

Niobium is recovered from scrap uranium-niobium alloy by melting the scrap with tin, solidifying the billet thus formed, heating the billet to combine niobium with tin therein, placing the billet in hydrochloric acid to dissolve the uranium and form a precipitate of niobium stannide, then separating the precipitate from the acid.

Case Study/ Effects of Groundwater Development on Uranium: Central Valley, California, USA Abstract Uranium (U) concentrations in groundwater in several parts of the eastern San Joaquin Valley development during the last 100 years have changed the chemistry and magnitude of groundwater recharge

Northwest Laboratory, Richland, Washington 99352 Uranium (U) solid-state speciation in vadose zone sediments of past nuclear fuel fabrication processes, uranium (U) has been recognized as one of the most widespreadHanfordsitesthatreceivedU-containingwastesduring its mission of Pu production between 1940 and 1990. Unirradiated fuel rod wastes were disposed

The report Uranium in US Surface, Ground, and Domestic Waters comprises four volumes. Volumes 2, 3, and 4 contain data characterizing the location, sampling date, type, use, and uranium conentrations of 89,994 individual samples presented in tabular form. The tabular data in volumes 2, 3, and 4 are summarized in volume 1 in narrative form and with maps and histograms.

An in-line assay monitor for determining the content of uranium-235 in a uranium hexafluoride gas isotopic separation system is provided which removes the necessity of complete access to the operating parameters of the system for determining the uranium-235 content. The monitor is intended for uses such as safeguard applications to assure that weapons grade uranium is not being produced in an enrichment cascade. The method and monitor for carrying out the method involve cooling of a radiation pervious chamber connected in fluid communication with the selected point in the system to withdraw a specimen and solidify the specimen in the chamber. The specimen is irradiated by means of an ionizing radiation source of energy different from that of the 185 keV gamma emissions from the uranium-235 present in the specimen. Simultaneously, the gamma emissions from the uranium-235 of the specimen and the source emissions transmitted through the sample are counted and stored in a multiple channel analyzer. The uranium-235 content of the specimen is determined from the comparison of the accumulated 185 keV energy counts and the reference energy counts. The latter is used to measure the total uranium isotopic content of the specimen. The process eliminates the necessity of knowing the system operating conditions and yet obtains the necessary data without need for large scintillation crystals and sophisticated mechanical designs.

Uranium-233 ({sup 233}U) is a synthetic isotope of uranium formed under reactor conditions during neutron capture by natural thorium ({sup 232}Th). At high purities, this synthetic isotope serves as a crucial reference material for accurately quantifying and characterizing uranium-bearing materials assays and isotopic distributions for domestic and international nuclear safeguards. Separated, high purity {sup 233}U is stored in vaults at Oak Ridge National Laboratory (ORNL). These materials represent a broad spectrum of {sup 233}U from the standpoint of isotopic purity - the purest being crucial for precise analyses in safeguarding uranium. All {sup 233}U at ORNL is currently scheduled to be disposed of by down-blending with depleted uranium beginning in 2015. This will reduce safety concerns and security costs associated with storage. Down-blending this material will permanently destroy its potential value as a certified reference material for use in uranium analyses. Furthermore, no credible options exist for replacing {sup 233}U due to the lack of operating production capability and the high cost of restarting currently shut down capabilities. A study was commissioned to determine the need for preserving high-purity {sup 233}U. This study looked at the current supply and the historical and continuing domestic need for this crucial isotope. It examined the gap in supplies and uses to meet domestic needs and extrapolated them in the context of international safeguards and security activities - superimposed on the recognition that existing supplies are being depleted while candidate replacement material is being prepared for disposal. This study found that the total worldwide need by this projection is at least 850 g of certified {sup 233}U reference material over the next 50 years. This amount also includes a strategic reserve. To meet this need, 18 individual items totaling 959 g of {sup 233}U were identified as candidates for establishing a lasting supply of certified reference materials (CRM), all having an isotopic purity of at least 99.4% {sup 233}U and including materials up to 99.996% purity. Current plans include rescuing the purest {sup 233}U materials during a 3-year project beginning in FY 2012 in three phases involving preparations, handling preserved materials, and cleanup. The first year will involve preparations for handling the rescued material for sampling, analysis, distribution, and storage. Such preparations involve modifying or developing work control documents and physical preparations in the laboratory, which include preparing space for new material-handling equipment and procuring and (in some cases) refurbishing equipment needed for handling {sup 233}U or qualifying candidate CRM. Once preparations are complete, an evaluation of readiness will be conducted by independent reviewers to verify that the equipment, work controls, and personnel are ready for operations involving handling radioactive materials with nuclear criticality safety as well as radiological control requirements. The material-handling phase will begin in FY 2013 and be completed early in FY 2014, as currently scheduled. Material handling involves retrieving candidate CRM items from the ORNL storage facility and shipping them to another laboratory at ORNL; receiving and handling rescued items at the laboratory (including any needed initial processing, acquisition and analysis of samples from each item, and preparation for shipment); and shipping bulk material to destination labs or to a yet-to-be-designated storage location. There are seven groups of {sup 233}U identified for handling based on isotopic purity that require the utmost care to prevent cross-contamination. The last phase, cleanup, also will be completed in 2014. It involves cleaning and removing the equipment and material-handling boxes and characterizing, documenting, and disposing of waste. As part of initial planning, the cost of rescuing candidate {sup 233}U items was estimated roughly. The annualized costs were found to be $1,228K in FY 2012, $1,375K in FY 2013,

Uranium mining and milling operations result in the release of radon from numerous sources of various types and strengths. The US Environmental Protection Agency (EPA) under the Clean Air Act, is assessing the health impact of air emissions of radon from underground uranium mines. In this case, the radon emissions may impact workers and residents in the mine vicinity. To aid in this assessment, the EPA needs to know how mine releases can affect the radon concentrations at populated locations. To obtain this type of information, Pacific Northwest Laboratory used the radon emissions, release characteristics and local meterological conditions for a number of mines to model incremental radon concentrations. Long-term, average, incremental radon concentrations were computed based on the best available information on release rates, plume rise parameters, number and locations of vents, and local dispersion climatology. Calculations are made for a model mine, individual mines, and multiple mines. Our approach was to start with a general case and then consider specific cases for comparison. A model underground uranium mine was used to provide definition of the order of magnitude of typical impacts. Then computations were made for specific mines using the best mine-specific information available for each mine. These case study results are expressed as predicted incremental radon concentration contours plotted on maps with local population data from a previous study. Finally, the effect of possible overlap of radon releases from nearby mines was studied by calculating cumulative radon concentrations for multiple mines in a region with many mines. The dispersion model, modeling assumptions, data sources, computational procedures, and results are documented in this report. 7 refs., 27 figs., 18 tabs.

The determination of the enrichment of uranium is required in many safeguards and security applications. Typical methods of determining the enrichment rely on detecting the 186 keV gamma ray emitted by {sup 235}U. In some applications, the uranium is surrounded by external shields, and removal of the shields is undesirable. In these situations, methods relying on the detection of the 186 keV gamma fail because the gamma ray is shielded easily. Oak Ridge National Laboratory (ORNL) has previously measured the enrichment of shielded uranium metal using active neutron interrogation. The method consists of measuring the time distribution of fast neutrons from induced fissions with large plastic scintillator detectors. To determine the enrichment, the measurements are compared to a calibration surface that is created from Monte Carlo simulations where the enrichment in the models is varied. In previous measurements, the geometry was always known. ORNL is extending this method to situations where the geometry and materials present are not known in advance. In the new method, the interrogating neutrons are both time and directionally tagged, and an array of small plastic scintillators measures the uncollided interrogating neutrons. Therefore, the attenuation through the item along many different paths is known. By applying image reconstruction techniques, an image of the item is created which shows the position-dependent attenuation. The image permits estimating the geometry and materials present, and these estimates are used as input for the Monte Carlo simulations. As before, simulations predict the time distribution of induced fission neutrons for different enrichments. Matching the measured time distribution to the closest prediction from the simulations provides an estimate of the enrichment. This presentation discusses the method and provides results from recent simulations that show the importance of knowing the geometry and materials from the imaging system.

Under the Memorandum of Understanding (MOU) between the EIA, US Department of Energy, and the US Geological Survey (USGS), US Department of the Interior, the USGS develops estimates of uranium endowment for selected geological environments in the United States. New estimates of endowment are used to update the Uranium Resources Assessment Data (URAD) System which, beginning in 1990, is maintained for EIA by the USGS. For 1989, estimates of US undiscovered resources were generated using revised economic index values (current to December 1989) in the URAD system's cost model. The increase in the estimates for the Estimated Additional Resources (EAR) and Speculative Resources (SR) classes resulted primarily from increases in the estimates of uranium endowment for the solution-collapse, breccia-pipe uranium deposit environment in the Colorado Plateau resource region. The mean values for $30-, $50-, and $100-per-pound U{sub 3}O{sub 8} forward-cost categories of EAR increased by about 8, 48, and 32 percent, respectively, as compared to 1988. Estimates of the 1989 undiscovered resources in the SR class also increased in all three forward-cost categories by 10, 5, and 9 percent, respectively. The original cost equations in the URAD System were designed to cover drilling costs related to extensive flat-lying tabular ore bodies. The equations do not adequately treat drilling costs for the smaller areas of vertical breccia pipe uranium deposits in the Colorado Plateau resource region. The development of appropriate cost equations for describing the economics of mining this type of deposit represents a major new task. 12 refs., 4 figs., 5 tabs.

Results of a study of the content of radium-226 in plants growing on inactive uranium mill tailings sites in the Four Corners Region of the southwestern United States and in plants grown under greenhouse conditions with minimal surficial contamination are reported. Field plant samples and associated substrates were analyzed from two carbonate tailings sites in the Grants Mineral Belt of New Mexico. Radium activities in air-cleaned samples ranged from 5 to 368 pCi/g (dry weight) depending on species and location: activities in plants growing on local soils averaged 1.0 pCi/g. The talings and local soils contain 140 to 1400 pCi/g and 2.1 pCi/g, respectively. An evaluation of cleaning methods on selected samples showed that from 17 to 79% of the radium activity measured in air-cleaned samples was due to surficial contamination, which varied with species and location. A survey of 18 inactive uranium mill sites in the Four Corners Region was performed. Radium activity in plant tissues from nine species ranged from 2 to 210 pCi/g on bare tailings and from 0.3 to 30 pCi/g on covered tailings The radium content in most of the soil overburdens on the covered tailings piles was 10 to 17 pCi/g. An experiment was performed to measure radium-226 uptake by two species grown on tailings covered with a shallow (5 cm) soil layer. A grass, Sporobolus airoides (alkali sacaton) and a shrub, Atriplex canescens (four-wing saltbush), were studied. The tailings were a mixture of sands and slimes from a carbonate pile. The tailings treatments were plants grown in a soil cover over tailings; the controls were plants grown only in soil. Three soil types, dune sand, clay loam, and loam, were used. The radium activity of the plant tissue from the tailings treatment compared to that of the appropriate control was 1 to 19 times greater for the grass and 4 to 27 times greater for the shrub.

The objective of the treatability test was to evaluate the efficacy of using polyphosphate injections to treat uranium-contaminated groundwater in situ. A test site consisting of an injection well and 15 monitoring wells was installed in the 300 Area near the process trenches that had previously received uranium-bearing effluents. This report summarizes the work on the polyphosphate injection project, including bench-scale laboratory studies, a field injection test, and the subsequent analysis and interpretation of the results. Previous laboratory tests have demonstrated that when a soluble form of polyphosphate is injected into uranium-bearing saturated porous media, immobilization of uranium occurs due to formation of an insoluble uranyl phosphate, autunite [Ca(UO2)2(PO4)2•nH2O]. These tests were conducted at conditions expected for the aquifer and used Hanford soils and groundwater containing very low concentrations of uranium (10-6 M). Because autunite sequesters uranium in the oxidized form U(VI) rather than forcing reduction to U(IV), the possibility of re-oxidation and subsequent re-mobilization is negated. Extensive testing demonstrated the very low solubility and slow dissolution kinetics of autunite. In addition to autunite, excess phosphorous may result in apatite mineral formation, which provides a long-term source of treatment capacity. Phosphate arrival response data indicate that, under site conditions, the polyphosphate amendment could be effectively distributed over a relatively large lateral extent, with wells located at a radial distance of 23 m (75 ft) reaching from between 40% and 60% of the injection concentration. Given these phosphate transport characteristics, direct treatment of uranium through the formation of uranyl-phosphate mineral phases (i.e., autunite) could likely be effectively implemented at full field scale. However, formation of calcium-phosphate mineral phases using the selected three-phase approach was problematic. Although amendment arrival response data indicate some degree of overlap between the reactive species and thus potential for the formation of calcium-phosphate mineral phases (i.e., apatite formation), the efficiency of this treatment approach was relatively poor. In general, uranium performance monitoring results support the hypothesis that limited long-term treatment capacity (i.e., apatite formation) was established during the injection test. Two separate overarching issues affect the efficacy of apatite remediation for uranium sequestration within the 300 Area: 1) the efficacy of apatite for sequestering uranium under the present geochemical and hydrodynamic conditions, and 2) the formation and emplacement of apatite via polyphosphate technology. In addition, the long-term stability of uranium sequestered via apatite is dependent on the chemical speciation of uranium, surface speciation of apatite, and the mechanism of retention, which is highly susceptible to dynamic geochemical conditions. It was expected that uranium sequestration in the presence of hydroxyapatite would occur by sorption and/or surface complexation until all surface sites have been depleted, but that the high carbonate concentrations in the 300 Area would act to inhibit the transformation of sorbed uranium to chernikovite and/or autunite. Adsorption of uranium by apatite was never considered a viable approach for in situ uranium sequestration in and of itself, because by definition, this is a reversible reaction. The efficacy of uranium sequestration by apatite assumes that the adsorbed uranium would subsequently convert to autunite, or other stable uranium phases. Because this appears to not be the case in the 300 Area aquifer, even in locations near the river, apatite may have limited efficacy for the retention and long-term immobilization of uranium at the 300 Area site..

Epigenetic uranium deposits with potential commercial value have been found in the lower part of the upper Eocene to lower Miocene Sespe Formation near Ojai, in Ventura County, California. This report describes the geological and geochemical setting of these deposits and postulates a model for their origin. Several uranium deposits are located on Superior Ridge, a topographic high about 3 miles long located just south of White Ledge Peak and 6 to 9 miles west of Ojai (Photo 1). A single uranium deposit on Laguna Ridge is located about 3 miles south of Superior Ridge, and was included with the Superior Ridge deposits in the White Ledge Peak district. A few small deposits are known to exist in other parts of Ventura County. A preliminary model for uranium mineralization in the Sespe Formation postulated that the organic material necessary for concentrating the uranium by chemical reduction or precipitation originated as terrestrial humic acid or humate.

The Clifton Quadrangle, Arizona and New Mexico, was evaluated to identify environments and delineate areas favorable for uranium deposits. The evaluation used criteria formulated for the National Uranium Resource Evaluation program. Evidence for the evaluation was based on surface studies, hydrogeochemical and stream-sediment reconnaissance, and aerial radiometric surveys. The quadrangle encompasses parts of three physiographic provinces: the Colorado Plateau, the transition zone, and the Basin and Range. The one environment determined, during the present study, to be favorable for uranium deposits is the Whitewater Creek member of the Cooney tuff, which is favorable for magmatic-hydrothermal uranium deposits on the west side of the Bursum caldera. No other areas were favorable for uranium deposits in sandstone, limestone, volcanogenic, igneous, or metamorphic environments. The subsurface is unevaluated because of lack of information, as are areas where access is a constraint.

The recent nuclear renaissance has motivated many countries, especially developing nations, to plan and build nuclear power reactors. However, domestic low enriched uranium demands may trigger nations to construct indigenous enrichment facilities, which could be redirected to fabricate high enriched uranium for nuclear weapons. The potential advantages of establishing multinational uranium enrichment sites are numerous including increased low enrichment uranium access with decreased nuclear proliferation risks. While multinational nuclear initiatives have been discussed, Russia is the first nation to actualize this concept with their Angarsk International Uranium Enrichment Center (IUEC). This paper provides an overview of the historical and modern context of the multinational nuclear fuel cycle as well as the evolution of Russia's IUEC, which exemplifies how international fuel cycle cooperation is an alternative to domestic facilities.

Utilizing amidoxime-based polymer sorbents for extraction of uranium from seawater has attracted considerable interest in recent years. Uranium collected in the sorbent is recovered typically by elution with an acid. One drawback of acid elution is deterioration of the sorbent which is a significant factor that limits the economic competitiveness of the amidoxime-based sorbent systems for sequestering uranium from seawater. Developing innovative elution processes to improve efficiency and to minimize loss of sorbent capacity become essential in order to make this technology economically feasible for large-scale industrial applications. This project has evaluated several elution processes including acid elution, carbonate elution, and supercritical fluid elution for recovering uranium from amidoxime-based polymer sorbents. The elution efficiency, durability and sorbent regeneration for repeated uranium adsorption- desorption cycles in simulated seawater have been studied. Spectroscopic techniques are used to evaluate chemical nature of the sorbent before and after elution. A sodium carbonate-hydrogen peroxide elution process for effective removal of uranium from amidoxime-based sorbent is developed. The cause of this sodium carbonate and hydrogen peroxide synergistic leaching of uranium from amidoxime-based sorbent is attributed to the formation of an extremely stable uranyl peroxo-carbonato complex. The efficiency of uranium elution by the carbonate-hydrogen peroxide method is comparable to that of the hydrochloric acid elution but damage to the sorbent material is much less for the former. The carbonate- hydrogen peroxide elution also does not need any elaborate step to regenerate the sorbent as those required for hydrochloric acid leaching. Several CO2-soluble ligands have been tested for extraction of uranium from the sorbent in supercritical fluid carbon dioxide. A mixture of hexafluoroacetylacetone and tri-n-butylphosphate shows the best result but uranium removal from the sorbent reaches only 80% after 10 hours of leaching. Some information regarding coordination of vanadium with amidoxime molecules and elution of vanadium from amidoxime- based sorbents is also given in the report.

The research in this thesis covers the design and implementation of a depleted uranium (DU) powder production system and the initial results of a DU-Zr-Mg alloy alpha phase sintering experiment where the Mg is a surrogate ...

An electrorefining process and apparatus for the recovery of uranium and a mixture of uranium and plutonium from spent fuel using an electrolytic cell having a lower molten cadmium pool containing spent nuclear fuel, an intermediate electrolyte pool, an anode basket containing spent fuel, and two cathodes, the first cathode composed of either a solid alloy or molten cadmium and the second cathode composed of molten cadmium. Using this cell, additional amounts of uranium and plutonium from the anode basket are dissolved in the lower molten cadmium pool, and then substantially pure uranium is electrolytically transported and deposited on the first alloy or molten cadmium cathode. Subsequently, a mixture of uranium and plutonium is electrotransported and deposited on the second molten cadmium cathode.

An electrorefining process and apparatus for the recovery of uranium and a mixture of uranium and plutonium from spent fuels is disclosed using an electrolytic cell having a lower molten cadmium pool containing spent nuclear fuel, an intermediate electrolyte pool, an anode basket containing spent fuels, two cathodes and electrical power means connected to the anode basket, cathodes and lower molten cadmium pool for providing electrical power to the cell. Using this cell, additional amounts of uranium and plutonium from the anode basket are dissolved in the lower molten cadmium pool, and then purified uranium is electrolytically transported and deposited on a first molten cadmium cathode. Subsequently, a mixture of uranium and plutonium is electrotransported and deposited on a second cathode. 3 figs.

Three unusual incidents and case histories involving uranium hexafluoride in the enrichment facilities of the USA in the late 1940`s and early 1950`s are presented. The history of the measurements of the thermal expansion of liquids containing fluorine atoms within the molecule is reviewed with special emphasis upon uranium hexafluoride. A comparison is made between fluorinated esters, fluorocarbons, and uranium hexafluoride. The quantitative relationship between the thermal expansion coefficient, a, of liquids and the critical temperature, T{sub c} is presented. Uranium hexafluoride has an a that is very high in a temperature range that is used by laboratory and production workers - much higher than any other liquid measured. This physical property of UF{sub 6} has resulted in accidents involving filling the UF{sub 6} containers too full and then heating with a resulting rupture of the container. Such an incident at a uranium gaseous diffusion plant is presented. Production workers seldom {open_quotes}see{close_quotes} uranium hexafluoride. The movement of UF{sub 6} from one container to another is usually trailed by weight, not sight. Even laboratory scientists seldom {open_quotes}see{close_quotes} solid or liquid UF{sub 6} and this can be a problem at times. This inability to {open_quotes}see{close_quotes} the UF{sub 6}-HF mixtures in the 61.2{degrees}C to 101{degrees}C temperature range caused a delay in the understanding of the phase diagram of UF{sub 6}-HF which has a liquid - liquid immiscible region that made the eutectic composition somewhat elusive. Transparent fluorothene tubes solved the problem both for the UF{sub 6}-HF phase diagram as well as the UF{sub 6}-HF-CIF{sub 3} phase diagram with a miscibility gap starting at 53{degrees}C. The historical background leading to the first use of CIF{sub 3} to produce UF{sub 6} in both the laboratory and plant at K-25 is presented.

Correlations between modifying electron donor and acceptor accessibility, the in-situ microbial community, and bioreduction of Uranium at the FRC and UMTRA research sites indicated that significant modifications in the rate, amount and by inference the potential stability of immobilized Uranium are feasible in these environments. The in-situ microbial community at these sites was assessed with a combination of lipid and real-time molecular techniques providing quantitative insights of effects of electron donor and manipulations. Increased (9mM in 2003 vs 3mM 2002) donor amendment at the Old Rifle site resulted in the stimulation of anaerobic conditions downgradient of the injection gallery. Biomass within the test plot increased relative to the control well at 17 feet. Q-PCR specific for IRB/SRB showed increased copy numbers within the test plot and was the highest at the injection gallery. Q-PCR specific for Geobacter sp. showed increased copy numbers within the test plot but further downgradient from the injection gallery than the SRB/IRB. DNA and Lipid analysis confirm changes in the microbial community structure due to donor addition. See also the PNNL (Long) and UMASS (Anderson) posters for more information about this site.

The United States Uranium Registry (USUR) was formed in 1978 to investigate potential hazards from occupational exposure to uranium and to assess the need for special health-related studies of uranium workers. This report provides a summary of Registry work done to date. The history of the uranium industry is outlined first, and the current commercial uranium industry (mining, milling, conversion, enrichment, and fuel fabrication) is described. This description includes information on basic processes and areas of greatest potential radiological exposure. In addition, inactive commercial facilities and other uranium operations are discussed. Regulation of the commercial production industry for uranium fuel is reported, including the historic development of regulations and the current regulatory agencies and procedures for each phase of the industry. A review of radiological health practices in the industry - facility monitoring, exposure control, exposure evaluation, and record-keeping - is presented. A discussion of the nonradiological hazards of the industry is provided, and the final section describes the tissue program developed as part of the Registry.

The separative elution of uranium from an amidoxime polymer was examined by the column method with hydrochloric acid solutions. The amidoxime polymer was immersed in seawater for 40 d for preparation of an uranium-loaded polymer sample for the elution experiments; the metal ions adsorbed were Mg(II), Ca(II), Fe(III), Ni(II), Cu(II), and Zn(II) as well as UO/sub 2/(VI). It was found from the pH dependence of elution extent by a batch method that the order of elution pH values is Fe(III) < UO/sub 2/(VI) < Cu(II) < Ni(II) < Zn(II) < Ca(II) < Mg(II). In the elution by a column method, Mg(II), Ca(II), Zn(II), and Ni(II) were eluted completely by 0.1 M HCl and the eluate of enriched uranium was obtained by a succeeding elution with 0.5 or 1 M HCl. This eluate contained Cu(II) and Fe(III), which could be removed in the succeeding step. The elution treatment with hydrochloric acid solutions hardly affected the adsorptivity for uranium in seawater. It was suggested that the elution of uranium with hydrochloric acid solutions from amidoxime polymers is satisfactorily applicable to uranium elution in the recovery of uranium from seawater with amidoxime polymers.

The Electrochemical Plant (ECP) is the one of the Russian Federation`s four uranium enrichment plants and one of three sites in Russia blending high enriched uranium (HEU) into commercial grade low enriched uranium. ECP is located approximately 200 km east of Krasnoyarsk in the closed city of Zelenogorsk (formerly Krasnoyarsk- 45). DOE`s MPC&A program first met with ECP in September of 1996. The six national laboratories participating in DOE`s Material Protection Control and Accounting program are cooperating with ECP to enhance the capabilities of the physical protection, access control, and nuclear material control and accounting systems. The MPC&A work at ECP is expected to be completed during fiscal year 2001.

The Urals Electrochemical Integrated Plant (UEIP) is the Russian Federation`s largest uranium enrichment plant and one of three sites in Russia blending high enriched uranium (HEU) into commercial grade low enriched uranium. UEIP is located approximately 70 km north of Yekaterinburg in the closed city of Novouralsk (formerly Sverdlovsk- 44). DOE`s MPC&A program first met with UEIP in June of 1996, however because of some contractual issues the work did not start until September of 1997. The six national laboratories participating in DOE`s Material Protection Control and Accounting program are cooperating with UEIP to enhance the capabilities of the physical protection, access control, and nuclear material control and accounting systems. The MPC&A work at UEIP is expected to be completed during fiscal year 2001.

by Galloway (1977) . Gal- loway's report concentrates on the "genetic stratigraphy, structural configuration, composition, and regional ground water flow dynamics of the Catahoula" and their relation to uranium mineralization. The following is a summary of... is actually located within a shaley sand and also a tuffaceous sand. In this study the minimum uranium concentration con- sidered to be ore was 50 ppm uranium. This value represents a significant enrichment in uranium over uranium concentra- tions found...

, physical and chemical subdivisions. The uranium-bearing solution is migrating from left to right, oxidizing the sand as it passes through. Uranium Acummulation Principal economic uranium deposits occur in terres- trial and shoreline sandstones... the primary mechan- ism responsible for uranium mineralization in south Texas. Most likely, none of' the above factors are solely respon- sible for the uranium mineralization of south Texas, but each factor probably plays a role in the development of...

A method for precipitating uranium from an aqueous solution and/or sediment comprising uranium and/or vanadium is presented. The method includes precipitating uranium as a uranyl vanadate through mixing an aqueous solution and/or sediment comprising uranium and/or vanadium and a solution comprising a monovalent or divalent cation to form the corresponding cation uranyl vanadate precipitate. The method also provides a pathway for extraction of uranium and vanadium from an aqueous solution and/or sediment.

This Conversion and Blending Facility (CBF) will have two missions: (1) convert HEU materials into pure HEU oxide and (2) blend the pure HEU oxide with depleted and natural uranium oxide to produce an LWR grade LEU product. The primary emphasis of this blending operation will be to destroy the weapons capability of large, surplus stockpiles of HEU. The blended LEU product can only be made weapons capable again by the uranium enrichment process. To the extent practical, the chemical and isotopic concentrations of blended LEU product will be held within the specifications required for LWR fuel. Such blended LEU product will be offered to the United States Enrichment Corporation (USEC) to be sold as feed material to the commercial nuclear industry. Otherwise, blended LEU will be produced as a waste suitable for storage or disposal.

Important advantages of heavy ion beam (HIB) irradiation of matter are discussed in comparison with traditional sources - laser heating, electron beam, electrical discharge etc. High penetration length (~ 10 mm) is of primary importance for investigation of dense matter properties. This gives an extraordinary chance to reach the uniform heating regime when HIB irradiation is being used for thermophysical property measurements. Advantages of HIB heating of highly-dispersive samples are claimed for providing free and relatively slow quasi-isobaric heating without fast hydrodynamic expansion of heated sample. Perspective of such HIB application are revised for resolution of long-time thermophysical problems for uranium and uranium-bearing compounds (UO2). The priorities in such HIB development are stressed: preferable energy levels, beam-time duration, beam focusing, deposition of the sample etc.

Electrochemical reprocessing is a technique used to chemically separate and dissolve the components of spent nuclear fuel, in order to produce new metal fuel. There are several different variations to electrochemical reprocessing. These variations are accounted for by both the production of different types of spent nuclear fuel, as well as different states and organizations doing research in the field. For this electrochemical reprocessing plant, the spent fuel will be in the metallurgical form, a product of fast breeder reactors, which are used in many nuclear power plants. The equipment line for this process is divided into two main categories, the fuel refining equipment and the fuel fabrication equipment. The fuel refining equipment is responsible for separating out the plutonium and uranium together, while getting rid of the minor transuranic elements and fission products. The fuel fabrication equipment will then convert this plutonium and uranium mixture into readily usable metal fuel.

The object of this thesis was to construct a temperature gradient furnace to simulate the thermal conditions in the reactor fuel and to study the migration of metallic inclusions in uranium oxide under the influence of temperature gradient. No thermal migration of molybdenum and tungsten inclusions was observed under the experimental conditions. Ruthenium inclusions, however, dissolved and diffused atomically through grain boundaries in slightly reduced uranium oxide. An intermetallic compound (probably URu/sub 3/) was formed by reaction of Ru and UO/sub 2-x/. The diffusivity and solubility of ruthenium in uranium oxide were measured.

substance and a corresponding enrichment in another. Soils, on being eroded, 14 adhorb dissolved uranium from runoff and ocean water and show a progressive change in U "/U activity ratios from 0. 9 in soils to 0, 95 in river muds to 1. 15 in recently... concluded that uranium is mainly associated with the non- carbonate fraction and reported his uranium and thorium concentra- tions on a calcium carbonate-free basis. His sediments were from the major oceans of the world; none were from the Gulf of Mexico...

pointing downd1p" (1). Th1s type of depos1t occurs when slightly basic, oxygenated, uranium-enriched ground water encounters an acidic reduc1ng layer of ground water, typ1cally conta1ning hydrogen sulfide, pyrite, and sandstone rich in organic matter... and organic matter has been recognized for many years. Berthoud (7), in 1875, , ' reported on the association of some uranium minerals with coal in , Colorado. In 1905, Boutwell (8) found coal1f1ed logs that also contained uranium minerals. In 1955...

Despite the low concentration of uranium in seawater (3.3 ppb), a special emphasis has been placed on its recovery. Although the concentration is low, it has been estimated that the world's oceans contain about 4 x 10[sup 9] tons of uranium - theoretically an unlimited supply of nuclear fuel. Adsorption has been considered to be a technically feasible procedure for a uranium recovery process with regard to economic and environmental impacts. The present paper restricts its coverage to those applications using chelating polymeric resins containing amidoxime groups as the most promising adsorbent. 72 refs., 8 figs., 1 tab.

The UO{sub 2} in spent nuclear fuel is unstable under moist oxidizing conditions and will be altered to uranyl oxide hydrate phases. The transuranics released during the corrosion of spent fuel may also be incorporated into the structures of secondary U{sup 6+} phases. The incorporation of radionuclides into alteration products will affect their mobility. A series of precipitation tests were conducted at either 150 or 90 C for seven days to determine the potential incorporation of Ce{sup 4+} and Nd{sup 3+} (surrogates for Pu{sup 4+} and Am{sup 3+}, respectively) into uranium phases. Ianthinite ([U{sub 2}{sup 4+}(UO{sub 2}){sub 4}O{sub 6}(OH){sub 4}(H{sub 2}O){sub 4}](H{sub 2}O){sub 5}) was produced by dissolving uranium oxyacetate in a solution containing copper acetate monohydrate as a reductant. The leachant used in these tests were doped with either 2.1 ppm cerium or 399 ppm neodymium. Inductively coupled plasma-mass spectrometer (ICP-MS) analysis of the solid phase reaction products which were dissolved in a HNO{sub 3} solution indicates that about 306 ppm Ce (K{sub d} = 146) was incorporated into ianthinite, while neodymium contents were much higher, being approximately 24,800 ppm (K{sub d} = 62). Solid phase examinations using an analytical transmission electron microscope/electron energy-loss spectrometer (AEM/EELS) indicate a uniform distribution of Nd, while Ce contents were below detection. Becquerelite (Ca[(UO{sub 2}){sub 6}O{sub 4}(OH){sub 6}]{center_dot}8H{sub 2}O) was produced by dissolving uranium oxyacetate in a solution containing calcium acetate. The leachant in these tests was doped with either 2.1 ppm cerium or 277 ppm neodymium. ICP-MS results indicate that about 33 ppm Ce (K{sub d}=16) was incorporated into becquerelite, while neodymium contents were higher, being approximately 1,300 ppm (K{sub d}=5). Homogeneous distribution of Nd in the solid phase was noted during AEM/EELS examination, and Ce contents were also below detection.

The hydrolysis reaction of uranium hexafluoride (UF{sub 6}) is a key step in the synthesis of uranium dioxide (UO{sub 2}) powder for nuclear fuels. Mechanisms for the hydrolysis reactions are studied here with density functional theory and the Stuttgart small-core scalar relativistic pseudopotential and associated basis set for uranium. The reaction of a single UF{sub 6} molecule with a water molecule in the gas phase has been previously predicted to proceed over a relatively sizeable barrier of 78.2 kJ {center_dot} mol{sup -1}, indicating this reaction is only feasible at elevated temperatures. Given the observed formation of a second morphology for the UO{sub 2} product coupled with the observations of rapid, spontaneous hydrolysis at ambient conditions, an alternate reaction pathway must exist. In the present work, two trimolecular hydrolysis mechanisms are studied with density functional theory: (1) the reaction between two UF{sub 6} molecules and one water molecule, and (2) the reaction of two water molecules with a single UF{sub 6} molecule. The predicted reaction of two UF{sub 6} molecules with one water molecule displays an interesting 'fluorine-shuttle' mechanism, a significant energy barrier of 69.0 kJ {center_dot} mol{sup -1} to the formation of UF{sub 5}OH, and an enthalpy of reaction ({Delta}H{sub 298}) of +17.9 kJ {center_dot} mol{sup -1}. The reaction of a single UF{sub 6} molecule with two water molecules displays a 'proton-shuttle' mechanism, and is more favorable, having a slightly lower computed energy barrier of 58.9 kJ {center_dot} mol{sup -1} and an exothermic enthalpy of reaction ({Delta}H{sub 298}) of -13.9 kJ {center_dot} mol{sup -1}. The exothermic nature of the overall UF{sub 6} + 2 {center_dot} H{sub 2}O trimolecular reaction and the lowering of the barrier height with respect to the bimolecular reaction are encouraging; however, the sizable energy barrier indicates further study of the UF{sub 6} hydrolysis reaction mechanism is warranted to resolve the remaining discrepancies between the predicted mechanisms and experimental observations.

Uranium-series data for groundwater samples from the Nopal I uranium ore deposit were obtained to place constraints on radionuclide transport and hydrologic processes for a nuclear waste repository located in fractured, unsaturated volcanic tuff. Decreasing uranium concentrations for wells drilled in 2003 are consistent with a simple physical mixing model that indicates that groundwater velocities are low ({approx}10 m/y). Uranium isotopic constraints, well productivities, and radon systematics also suggest limited groundwater mixing and slow flow in the saturated zone. Uranium isotopic systematics for seepage water collected in the mine adit show a spatial dependence which is consistent with longer water-rock interaction times and higher uranium dissolution inputs at the front adit where the deposit is located. Uranium-series disequilibria measurements for mostly unsaturated zone samples indicate that {sup 230}Th/{sup 238}U activity ratios range from 0.005-0.48 and {sup 226}Ra/{sup 238}U activity ratios range from 0.006-113. {sup 239}Pu/{sup 238}U mass ratios for the saturated zone are <2 x 10{sup -14}, and Pu mobility in the saturated zone is >1000 times lower than the U mobility. Saturated zone mobility decreases in the order {sup 238}U{approx}{sup 226}Ra > {sup 230}Th{approx}{sup 239}Pu. Radium and thorium appear to have higher mobility in the unsaturated zone based on U-series data from fractures and seepage water near the deposit.

Uranium urinalysis and in vivo examination results obtained from workers at eleven uranium mills between 1978 and 1980 were evaluated. The main purpose was to determine the degree of the mills' compliance with bioassay monitoring recommendations given in the draft NRC Regulatory Guide 8.22 (USNRC 1978). The effect of anticipated changes in the draft regulatory guidance, as expressed to PNL in May 1982, was also studied. Statistical analyses of the data showed that the bioassay results did not reliably meet the limited performance criteria given in the draft regulatory guide. Furthermore, quality control measurements of uranium in urine indicated that detection limits at ..cap alpha.. = ..beta.. = 0.05 ranged from 13 ..mu..g/l to 29 ..mu..g/l, whereas the draft regulatory guidance suggests 5 ..mu..g/l as the detection limit. Recommendations for monitoring frequencies given in the draft guide were not followed consistently from mill to mill. The results of these statistical analyses indicate a need to include performance criteria for accuracy, precision, and confidence in revisions of the draft Regulatory Guide 8.22. Revised guidance should also emphasize the need for each mill to continually test the laboratory performing urinalyses by submitting quality control samples (i.e., blank and spiked urine samples as open and blind test) to insure that the performance criteria are being met. Recommendations for a bioassay audit program are also given. 25 references, 15 figures, 17 tables.

In October 1978, Mesa College entered into subcontract with Bendix Field Engineering Corporation (BFEC) to prepare a bibliographic index of the uranium raw materials reports issued by the Grand Junction Office of the US Department of Energy (DOE). Bendix, prime contractor to the Grand Junction Office, operates the Technical Library at the DOE facility. Since the early 1950s, approximately 2700 reports have been issued by the Grand Junction Office. These reports were the results of uranium investigations conducted by federal agencies and their subcontractors. The majority of the reports cover geology, mineralogy, and metallurgy of uranium and/or thorium. No single, complete list of these reports existed. The purpose of this subcontract was to compile a comprehensive index to these reports. The Mesa College geology faculty worked with the BFEC and DOE staffs to develop the format for the index. Undergraduate geology students from Mesa compiled a master record sheet for each report. All reports issued up to January 1, 1979 were included in the bibliography. The bibliography is in preliminary, unedited form. It is being open-filed at this time, on microfiche, to make the information available to the public on a timely basis. The bibliography is divided into a master record list arranged in alpha-numeric order by report identification number, with separate indices arranged by title, author, state and county, 1/sup 0/ x 2/sup 0/ NTMS quadrangle, key words, and exploration area.

Every 30 to 40 million operating hours a destructive reaction is observed in one of the {approximately}4000 large compressors that move UF{sub 6} through the gaseous diffusion plants. Despite its infrequency, such a reaction can be costly in terms of equipment and time. Laboratory experiments reveal that the presence of moderate pressures of UF{sub 6} actually cools heated aluminum, although thermodynamic calculations indicate the potential for a 3000-4000{degrees}C temperature rise. Within a narrow and rather low (<100 torr; 1 torr = 133.322 Pa) pressure range, however, the aluminum is seen to react with sufficient heat release to soften an alumina boat. Three things must occur in order for aluminum to react vigorously with either UF{sub 6} or UO{sub 2}F{sub 2}. 1. An initiating source of heat must be provided. In the compressors, this source can be friction, permitted by disruption of the balance of the large rotating part or by creep of the aluminum during a high-temperature treatment. In the absence of this heat source, compressors have operated for 40 years in UF{sub 6} without significant reaction. 2. The film protecting the aluminum must be breached. Melting (of UF{sub 5} at 620 K or aluminum at 930 K) can cause such a breach in laboratory experiments. In contrast, holding Al samples in UF{sub 6} at 870 K for several hours produces only moderate reaction. Rubbing in the cascade can undoubtedly breach the protective film. 3. Reaction products must not build up and smother the reaction. While uranium products tend to dissolve or dissipate in molten aluminum, AIF{sub 3} shows a remarkable tendency to surround and hence protect even molten aluminum. Hence the initial temperature rise must be rapid and sufficient to move reactants into a temperature region in which products are removed from the reaction site.

A series of uranium-, thorium-, and gold-bearing conglomerates in Late Archean and Early Proterozoic metasedimentary rocks have been discovered in southern Wyoming. The mineral deposits were found by applying the time and strata bound model for the origin of uranium-bearing quartz-pebble conglomerates to favorable rock types within a geologic terrane known from prior regional mapping. No mineral deposits have been discovered that are of current (1981) economic interest, but preliminary resource estimates indicate that over 3418 tons of uranium and over 1996 tons of thorium are present in the Medicine Bow Mountains and that over 440 tons of uranium and 6350 tons of thorium are present in Sierra Madre. Sampling has been inadequate to determine gold resources. High grade uranium deposits have not been detected by work to date but local beds of uranium-bearing conglomerate contain as much as 1380 ppM uranium over a thickness of 0.65 meters. This project has involved geologic mapping at scales from 1/6000 to 1/50,000 detailed sampling, and the evaluation of 48 diamond drill holes, but the area is too large to fully establish the economic potential with the present information. This first volume summarizes the geologic setting and geologic and geochemical characteristics of the uranium-bearing conglomerates. Volume 2 contains supporting geochemical data, lithologic logs from 48 drill holes in Precambrian rocks, and drill site geologic maps and cross-sections from most of the holes. Volume 3 is a geostatistical resource estimate of uranium and thorium in quartz-pebble conglomerates.

The Uranium Mill Tailings Remedial Action (UMTRA) Project consists of the Surface Project (phase I), and the Ground Water Project (phase II). For the UMTRA Project site located near Naturita, Colorado (the Naturita site), phase I involves the removal of radioactively contaminated soils and materials and their transportation to a disposal site at Union Carbide Corporation`s Upper Burbank Repository at Uravan, Colorado, about 13 road miles (mi) (21 kilometers [km]) to the northwest. No uranium mill tailings are involved because the tailings were removed from the Naturita site and placed at Coke Oven, Colorado, during 1977 to 1979. Phase II of the project will evaluate the nature and extent of ground water contamination resulting from uranium processing and its effect on human health or the environment; and will determine site-specific ground water compliance strategies in accordance with the US Environmental Protection Agency (EPA) ground water standards established for the UMTRA Project. Human health risks could occur from drinking water pumped from a hypothetical well drilled in the contaminated ground water area. Environmental risks may result if plants or animals are exposed to contaminated ground water, or surface water that has received contaminated ground water. Therefore, a risk assessment is conducted for the Naturita site. This risk assessment report is the first site-specific document prepared for the Ground Water Project at the Naturita site. What follows is an evaluation of current and possible future impacts to the public and the environment from exposure to contaminated ground water. The results of this evaluation and further site characterization will be used to determine whether any action is needed to protect human health or the environment.

Uranium dioxide (UO2) is the major nuclear fuel component of fission power reactors. A key concern during severe accidents is the melting and leakage of radioactive UO2 as it corrodes through its zirconium cladding and steel containment. Yet, the very high temperatures (>3140 kelvin) and chemical reactivity of molten UO2 have prevented structural studies. In this work, we combine laser heating, sample levitation, and synchrotron x-rays to obtain pair distribution function measurements of hot solid and molten UO2. The hot solid shows a substantial increase in oxygen disorder around the lambda transition (2670 K) but negligible U-O coordination change. On melting, the average U-O coordination drops from 8 to 6.7 ± 0.5. Molecular dynamics models refined to this structure predict higher U-U mobility than 8-coordinated melts.